ML20137H635
| ML20137H635 | |
| Person / Time | |
|---|---|
| Site: | Comanche Peak  |
| Issue date: | 12/13/1982 |
| From: | Williams H TEXAS UTILITIES ELECTRIC CO. (TU ELECTRIC) |
| To: | |
| Shared Package | |
| ML17198A292 | List:
|
| References | |
| FOIA-85-59 QI-QP-11.0-3, NUDOCS 8512020436 | |
| Download: ML20137H635 (7) | |
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i INSTRUCTION l REVIS!CN ISSUE PAGE TEXAS UTILITIES GENERATING CO.
NUMBER i
CATE CPSES
,QI-QP-11.0-3 4
lDEC 13 1982i 1 of 7 I
PREPARED BY: / w/8,4
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T N T ON APPROVED BY:
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APPROVED BY:
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1.0 REFERENCES
i 1-A Specification 2323-SS-9, " Concrete" Y
h 1-B CEI-29, " Removable Concrete Blocks" 2.0 GENERAL I
2.1 PURPOSE AND SCOPE I
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To delineate the methods used by QC Inspectors to assure that C
concrete, mortar and concrete block placement activities are accomplished in accordance with Reference 1-A and 1-B.
3.0 INSTRUCTION 3.1 PRE-POUR PLANNING l
The QC Inspector will verify that the pre-planning provisions.
of Regulatory Guide 1.55 are being complied with.
This specifically means a demonstrated awareness that items such as cleanliness, availability of personnel and placing I
equipment, placing sequence and alternate construction joint locations have been provided for.
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On containment wall and dome placements, or those considered i
complex by the Civil Inspection Supervisor, the planning shall I
be in written form.
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3.2 INSPECTION REQUIREMENTS I
3.2.1 Prior to Placement l
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l Immediately prior to placement, the QC Inspector shall verify the following items:
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{1 INSTRUCTION ISSUE REVISION PAGE TEXAS UTILITIES GENERATING CO.
DME CPSES QI-QP-11.0-3 4
DEC 13 1982 2 of 7 a.
Pour card shall consist of a unique " Pour Number" for each concrete placement and signed by the craft and engineers, b.
Formwork for stability, line and grade, tightness of joints and cleanliness.
c.
The fabrication tolerance of concrete blocks is ! 1/8";
however, the overall depth of blocks in an opening shall not exceed the "As-Built" depth of the opening.
d.
Dimensions of seismic air gap are in compliance with drawings, and gap is protected from intrusion of foreign
- objects, e.
Proper installation of expanded metal construction joints.
f.
Construction joints clean and thoroughly wetted with no OL standing water, g.
Adequacy of placing personnel and equipment.
3.2.2 During Placement During placement, the QC Inspector shall verify the following items:
a.
Placing sequence conforms to the pre-pour plan.
b.
Placement rated ensures proper integration with preceding lift.
c.
Placemement lifts do not exceed 18".
d.
Placement technique causes minimum segregation, e.
Vibrators not used to convey concrete or mortor more than three feet.
f.
Vibrators penetrated approximately four inches into preceeding lift.
g.
Vibrators operate at minimum 8000 I.P.M. (Document on inspection report).
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Record four placement temperature measurements on l
inspection report.
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INSTRUCTION ISSUE
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NUMBER DATE RNSION AGE TEXAS UTILITIES GENERATING CO.
CPSES QI-QP-11.0-3 4
DEC 1. 3 1982 3 of 7 1.
Trucks discharged before 45 minutes if ambient temperature above 85'F., or 90* minutes if approved by Engineer.
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Placement does not contact aluminum.
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Placememnt placed to proper grade.
1.
Application of required finish.
3.3 REMOVABLE CONCRETE BLOCK INSTALLATION The QC Inspector shall monitor the concrete block installation, as per the QC Supervisor, in accordance with
" Typical Wall Concrete Block Arrangement" unless otherwise noted on Gibbs & Hill drawings.
The concrete blocks shall be staggered to meet shielding requirements per the applicable Gibbs & Hill drawings.
(O the oc t sPecete seeii me iter the em8er4 9 er the ce crete blocks.
Prior to concrete block installation all concrete blocks shall be numbered as per Attachment 2 using a Permanent Nuclear Marker (Marsh AEC) to maintain exact location and orientation of blocks for future removal.
i Rebar (main wall reinforcement) may be cut only for the installation of 1/2" $ thru bolts as shown on applicable-Gibbs & Hill drawings.
3.4 DOCUMENTATION OF INSPECTION Verify inspection to provisions of this instruction on the Inspection Reports (Attachments 1 and 3) and upon completion submit to the Civil Inspection Supervisor / designee for review, processing and filing in accordance with CPSES requirements for "QA Records".
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DEC 131982 OI-0P-11.0-3 4 of 7 ATTACHMENT 1 CCa4 AN ChE PCAE ST DAG E!.!CTRIC S TATION y
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'O' ANTS AND CLEANLINESS. J0!NT5 GROUT TIGHT. PARA 3.2.1 II l
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! PROPER INSTALLATION OF EXPANDED METAL CONSTRUCTION ll l
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IPLACEMENT RATE ENSURES PROPER INTFGRATION WITH PRECEDING Il l
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INSTRUCTION ISSUE REVISION AGE NUMBER DATE TEXAS UTILITIES GENERATING CO.
]I-QP-11.0-3 4
ATTACHMENT 1 (CONT.)
CDMANCHE PEAK STEAM ELECTRIC STATION (supp J M m )
Sheet 2 of 2_
l FOR FULL HEADINGS SEE SHEET 1
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INSM:CTION ATTRieuTES 5
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,Q, INSTRUCTION ISSUE REVISION PAGE NUMBER DATE TEXAS UTIUTIES GENERATING CO.
CPSES DEC 13 1982 7 of 7 OI-QP-11.0- 3 4
ATTACHMENT 1 COM AN CHE PEAK STEAM ELECTRIC S TATION wtr' or INSPECTION REPORT lan ertu ocacnierio" f1DNITOR OF CONCRET('80"""' Aran a prmm/sinwCw maarica
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2658 April 25, 1983 h
APPENDIX A
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NOTICE OF VIOLATION Texas Utilities Generating Company Docket:
50-445/83-03 r
Comanche Peak Steam Electric Station 50-446/83-01 Permits: CPPR-126 CPPR-127 Based on the results of an NRC inspection conducted during the period of October 1982 through February 1983, and in accordance with the NRC Enforcement Policy (10 CFR Part 2, Appendix C), 47 FR 9987, dated March 9,1982, the following violation was identified:
Failure to Imolement a Quality Assurance Program for the Fabrication and Installation of Electrical Underwater Floodlight Pole Assemolies Criterion II of Appendix B to 10 CFR 50 requires that the applicant shall identify the structures, systems, and components to be covered by the quality assurance program and that the program shall provide control over activities affecting quality of the identified structures, systems and components.
FSAR Section 1A(b) commits the applicant to '
compliance with NRC Regulatory Guide 1.29 which in paragraphs 2 and 4
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require the applicant to identify those structures, systems, and components whose continued function is not required (in a design basis accident) but whose failure could reduce the functioning of any plant feature identified in other paragraphs to an unacceptable level.
Contrary to the above, the Senior Resident Inspector-Construction has determined from investigation of allegations, observation of construction activities and review of design drawings that group of devices collectively identified as " Electrical Underwater Floodlighting Poles" (Drawing 2323-EL-0925-02) were not identified as required by Regulatory Guide 1.29 and were not included within the licensee's Quality Assurance Program. Mechanical failure of the devices in a seismic event could damage fuel during reactor core installation activities or in the spent fuel storage pools, although the possibility of such mechanical failure of the pole assembly resulting in damaging fuel is very remote due to the design of upper and lower pole retention devices.
This is a Severity Level V Violation.
(Supplement II.D.)
CORRECTIVE STEPS WHICH HAVE BEEN TAKEH AND THE RESULTS ACH~EVED_
The drawing governing installation of the light poles has been : hanged vio I
Design Change Authorization to reflect Seismic Categcry II requirements.
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-2 April 25, 1983 change causes the light poles to be inspected in accordance with the CPSES QA Program. Any discrepancies discovered have been or will be documented and resolved in accordance with established procedures.
CORRECTIVE STEPS WHICH HAVE BEEN OR WILL BE TAKEN TO AVOID FURTHER VIOLATIONS Items designat.ed "Non-Nuclear Safety Related" have been previously reviewed for applicable Seismic Category II requirements and upgraded as needed.
Considering the scope of the evaluation and subsequent results, adequate confidence has been established that appropriate QA Program controls are in effect.
DATE OF FULL COMPLIANCE The required drawing change was accomplished on February 4,1983.
Inspection activities are on-going and will be completed and resolved consistent with the project schedule.
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AUG 2 4 1983 In Reply Refer To:
Docket: 50-445/S3-24
/ Y f 50-446/83-15
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/f Texas Utilities Generating Company ATTN:
R. J. Gary Executive Vice Pmsident & General Manager 2001 Bryan Tower j g /7 Dallas. Texas 75201 o
5 Gentlemen:
This refers to the inspection conducted by our Senior Resident Inspector, Construction, Mr. R. G. Taylor, during the period March through July 1983, s
of activities authorized bf NRC Construction Permits CPPR-126 and CPPR-127 for Comanche Peak, Units 1 and 2, and to the discussion of our findings with
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Mr. R. G. Tolson, and other members of your staff during the inspection.
- c. Areas examined during the inspection included reviewisinspection, and evalua-tion ofleveral allegations made to various NRC persdns, including the Atonde Safety and Licensing Board in their proceedings regarding the operating license for Cananche Peak Steam Electric Station (CPSES).
Within these areas, the inspection consisted of selective examination of procedures and representative records, interviews with personnel, and observations by the inspector. These
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findings are documented in the enclosed inspection report.
During this inspection, it was found that certain of your activities were in violation with NRC requimments. You were notified of one such violation by our letter of May 31, 1983, to which you have responded.
Details of the 1
item enclosed with our My 31, 1983 letter are included in the enclosed l
inspection report.
i unresolved item is identified in paragraph 15 of the enclosed inspection
, e have also examined actions you have taken with regard to previously W
identified inspection findings.
The status of these items is identified l
j in paragraph 2 of the enclosed mport.
In accordance with 10 CFR 2.790(a), a copy of this letter and the enclosure will be placed in the NRC Public Document Roam unless you notify this office, by telephone, within 10 days of the date of this letter, and submit written application to withhold infomation contained therein within 30 days of the date of this letter.
Such application must be consistent with the require-ments of 2.790(b)
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2 Ce:::pany Aug g Should you have any questions concerning this inspection, we will be pleased to ciscuss them with you.
Sincerely.
-onsinni staned
- C. L. MADcW G. L. t'adsen, Chief Reactor Project Branch 1
Enclosure:
Appendix - NRC Inspection Report 50-445/83-24 50-446/53-15 cc w/encls:
Texas Utilities Generating Company ATTH: H. C. Schmidt, Project Panager 2001 Bryan Tower Callas, Texas 75201
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ATTH: B. R. Clements, Vice President, Nuclear 2001 Bryan Tower, Suite 1735 rellas, Texas 75201 bec to Dts (IE01) bec distrib. by RIV: '
RPB1 D. Kelley, SRI-Ops RPS2 R. Taylor, SRI-Cons ~
TPB Section Chief (RPS-A)
J. Collins, RA J. Gagliardo, DRRP&EP C. Wisner, PA0 M. Rothschild, ELD MIS SYSTEM RIV File TEXAS STATE DEPT. OF HEALTH Juanita Ellis David Preister 4
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N APPENDIX U. S. NUCLEAR REGULATORY COMMISSION REGION IV r
NRC Inspection Report: 50-445/S3-24 50-446/83-15 Docket: 50-445.
Ca tegory:
A2 50-446 Licensee:
Texas Utilities Generating Company (TUGCO) 2001 Bryan Tower Dallas, Texas, 75201 Facility Name:
Comanche Peak Steam Electric Station (CPSES), Units 1 and 2 Inspection At:
Comanche Peak, Units 1 and 2, Glen Rose Texas Inspection Conducted: March through July 1983 or Inspectors:
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83 R. G. Taylor, Senior P.esident Inspector Date /
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Construction (SRIC) h)& M 8//f/83 Approved:
D. M. Hunnicutt, Chief.
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Reactor Pro, ject Section A i
Inspection Sumary Inspection Conducted March through July 1983 (Report 50-445/83-24 and 83-446/83-15)
Areas Insoected: Special inspections, announced and unannounced, related to allegations made to various NRC persons including the Atomic Safety and Licensing Board in their procedings regarding the operating license for Comanche Peak Station.
The inspections involved 449 inspector-hours by one NRC inspector.
Results: The inspection confirmed the need to issue four violations initially icentified by the Construction Appraisal Team (CAT) (NRC Inspection Report 50-445/83-18; 50-446/83-12).
These involved the areas of HVAC, Equipment Installation, Document Control, and Storage of Eouipment! ~
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P_ersons Contacted Princioal Licensee Employees i
- R. G. Tolson, Site QA Supervisor
- C. T. Brandt, Non-ASME QC Supervisor
- J. R. Merritt, Engineering, Construction and Startup Manager
- J. B. George, Project General Manger
- D. N. Chapman, QA Manager
- B. R. Clements, Vice-President, Nuclear Brown & Root (B&R)
- G. R. Purdy, Project QA Manager
- D. Frankum, Construction Project Manager The SRIC also ' interviewed many other licensee, B&R, and subcontractor l ~
persor.nel during the course of the inspectionr.
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- Denotes those persons who attended one or more management interviews with the SRIC.
2.
Licensee Action on Previous Inspection Findinos (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 r'estraint essentially surrounding the mainsteam and feed water lines located several feet outside of the ASME code boundry point.
The device was engineered by the licensee's A/E and manufactured by NPS Industries.
Due to the overall size of the structure, it has been nick-named " George Washington Bridge" by the site lebor and quality forces.
The licensee had reported the finding of the defect; as a potential 50.55(e) item to the SRIC on September 30, 1982, which w.s subsequently stated not reportable in a letter dated December 27, 1982.
An NRC inspector followed up on the matter during a visit to the offices of the A/E, as documented in NRC Inspection Report 50-445/83-12.
This review pertained to all of the defects involved with the exception of two cracked welds that had not been analyzed at the time of the inspection.
The engineer has recently analyzed these two defects and has determined that had they not been detected, the i
structure could have fulfilled it's function.
The SRIC has reviewed the location of the cracks and their length in relation to the size of the welds and the functional application of the structure.
Since the structure has no l
continuous service application and is essentially subject to a one-time loading, the cracks would not have the potential for further propagation.
j Further, the cracks are at points in the structure that would receive rela-tively low stresses in the one-time impact based on their small size in relation to the members being welded.
It appears that the cracks formed due to the stresses developed during the tightening of high strength bolting in
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the immediate vicinity of the welds during the site assembly of the structure.
Taken in conjunction with the earlier documented review of the engineers calculations and the SRIC's review of these cracks, the SRIC has concluded i
that the engineer's o'verall 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.
1 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 3
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) i The SRIC has reviewed a report of the licensee's self-evaluation performed j
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 lican-see by personnel in the employment of Sargent & Lundy, an architect-engineer firm with substantial nuclear power involvement.
A copy of the report was furnished to,the NRC, and subsequently, to tg Atomic Safety and Licensing Board in the matter of Comanche Peak Station Cpe~ rating license by letter dated May 2, 1983.
The purpose of the review by the SRIC was to detennine 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 recuired by 10 CFR 50.55(e).
The SRIC reviewed each of the 47 findings and the supporting documentation in the report pertaining to each finding.. This review revealed that none of the 47 items were based upon identified deficiencies in structures, systems, or components nor were there any significant deficiencies in design, 1
engineering,(e). testing that would constitute conditions reportable under 1
or 10 CFR 50.55 4.
Car Wash In Containinent During the limited appearance statement portion of the Atomic Safety and Licensin~g Board hearing on May 16, 1983, a person stated at transcript page 6152 that he understood that the containment looked something like a car wash.
The person stated that it was his understanding that the situa-tion developed at about the same time that there was a meeting at the D/FW Airport between the NRC and any interested' parties to discuss NRC decen-tralization.
That meeting took place on April 5,1983.. For the purposes of evaluating this allegation, the SRIC expanded the period of interest to include the 3 weeks prior to the meeting.
During this entire period, the Unit I reactor system was undergoing what is referred to as " Hot Func-tional Testing".
This particular test is an accurate simulation of the operation of the reactor system and its appurtenances but without a reactor core being in place.
The heat and pressure in the system is generated by the reactor coolant pumps in conjunction with the chemical and volume con-trol system charging pumps.
The test could readily be construed to be a pressure test but in fact is an operational test at pressure. This parti-cular test extended overall for about 90 days beginning lat'e in February
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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 rcom 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 dripping from the reactor coolant. pump motor cooling coils.
This was caused by the cold water in the coils condensing the humidity from the atmosphere within 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 r
vent valves were partially opened, which.in turn wo.uld give some amount of steam blowoff into the reactor refueling cavity area and would rise up into the building until cooled and c,ondensed out as. water.
None of these events rre typical of any major leak indicative of $PHng -or piping component (such as a valve) failure.
The type of smal1~ 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 perfonning the test.
5.
Desion of the HVAC System Supports By letters, both. dated March 11, 1983, Citizens Association for Sound Energy (CASE) notified the NRC's-Offices of Inspection and Enforcement and the Executive Legal Director of 4, concern that the HVAC system for Comanche i
Peak had not been properli' supported, nor had TT been properly consioered-
- in r'egard to seismic load c6hdiion. Or its treatment as potential mis-s il es'.
CASE specifically states that from their review of the FSAR, it appears that the licensee has not analyzed the HVAC supports for a 4,
seismic load condition.
Specific reference is made to Sheet 21 of Table 17A.
In addition, the personal observations of Messrs. Walsh and Doyle are l
relied upon to point out that there are no lateral suoDorts on the HVAC systems witnin the containment.~ CASE also states that all HVAC components and supports inside containment should be treated as missiles under Cri-terion 4 of the General Design Criteria for Nuclear Power Plants, 10 CFR 50, Appendix A.
Sheet 21 of Table 17A of the FSAR lists the containment ventilation sys-tems as being Seismic Category II. Apparently, it has been assumed by CASE that this category excludes seismic loading in the design.
This assumption is incorrect since the FSAR, Section 3.2.1.2 defines Seismic Category II a.s being those portions of systems or components whose j
i 4
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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 Category II and are designed and constructed so that a safe shutdown earthquake (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 which could also mean that it might be impossible for workers to enter the containment due to the heat.
Containment heat renoval 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 gf 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-et, asstaning the availability of either onsise.or.offsite power.: exclusively, 4
l does not prevent the system from accomplishing its planned safetPfunction.
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 detemined that all permanent HVAC supports are analyzed for seismic loading.
Two methods are utilized:
Zero Peak Accleration (IPA), or 1.5 Timest the Peak Accelera-tion When the Fundamental Frequency Falls Below 20 Hertz.
Of tne latter method of design, only about 6 out of 4000 supports have been designed that 4
way.
A typical HVAC duct run is supported axially at every third support This may explain why Messrs. Walsh and Doyle may have felt that there were no lateral supports on the HVAC systems.
The NRC inspector reviewed the design of a typical HVAC duct run at elevation 852'-6" in the Auxiliary Building.
Supports were designed utilizing two computer programs entitled i
FEASA-2D and FEASA-3D.
The acronym stands for frame eigenvalue and stress analysis.
The -2D version is used on the transverse supports and the -3D version is used on the axial supports.
The inclusion of equivalent weights
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from 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 utilized for the design of HVAC supports at Comanche Peak.
The concerns expressed by CASE have been found to be without merit.
I Ffersons contacted during the course of the inspection at Corporate Consulting e
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& Development Company, LTr. were:
J. Roland Yow, President & Cnief Executive Officer Gary Hughes, Vice-President for Operations David Lindley, Principal Engineer Stephen Lehrman, Seismic Department Manager Daryl Hughes, Project Engineer 6.
Heatino, 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 hinch leg size. The SRIC noted by visual comparison to the inch thick base metal that very few of the welds were of proper size.
The CAT inspectors also found cases where the bolting and gaskets between ducting sections were loose and/or missing..
i The CAT inspectors also found that some supp@t-members were noti within the dimensional tolerances on the design drawings.
It wai-noted that the contractor's inspection records did not reveal these various facts, indicating ineffectual QC by the contractor.
- Further, a review of the licensee's audit program indicated that the licensee was unaware of these several problems 11n the fabrication, installation, and inspection of the HVAC systems.
Based upon the CAT inspectors' findings and his,own observations, the SRIC recommended that a notice of violation be issued to the licensee pertaining collectively to these matters (Notice of Violation issued on Maf 31,1983.
I Reference 50-445/83-18 and 50-446/83-12, item 4).
1 7.
Installation of Major Itehs of Ecuipment l
The CAT inspectors noted during their inspections of certain major I'
items of equipment that there were several variables in how the equipment was fastened to the building equipment pads.
In some l'
instances, tanks for example, CAT inspectors found that there were two nuts (double nuts) on the embedded bolts securing the equipment, other bolts had one nut. (single nut) and some had a combination of s
both single nuts and double. nuts on one piece of equipment.
The
- CAT personnel also noted that certain heat exchangers had slotted holes in one of the mounting bases to allow for thermal expansion during operation. The holddown nuts appeared to be installed too tightly and may have prevented freedom of movement.
The SRIC obtained the design and installation drawings for two of the referenced heat exchangers identified in the CAT report.
Both were found to be horizontaT Utube heat exchangers whose function is nonsafety, but whose pressure boundary in the tubes'is safety-related since the process fluid could be radioactive. The SRIC found that the construction drawings for the mounting pedestals had a. flat steel plate on one m
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9 s %-2A-7 pedestal that would be suitable for the type of mounting detail on these heat exchangers.
The SRIC then reviewed the installation travelers for each heat exchanger and found that these documents did not note or address the slotted details, the plate, or the fact the bolts should be left loo:e.. The SRIC would note that the vendor tranual 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 reviev, 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 provisiens made to lock nuts where they must be deliberately left loose. *fMs was considered.
I 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 Equipment In Outdoor Storace Areas The CAT found that a considerable amount of equipment such as p'ipe support struts, clamps, and like items, normally stored. outdoors, was not being properly maintained in accordance with procedure MCP-10,
" Storage and Storage Maintenance of Mechanical and Electrical Equipment", as evidenced by rusting bolts and adjustment screws on s truts.
In addition, the strut bearings were dirty from dust and h
the bearing load pins, in some instances, were rusted.
By a tour
, of the storage areas, the SRIC confirmed the CAT inspectors find-p ings.
The SRIC would also note that the INPO Self-Evaluation S
Report at page 111 describes essentially the same finding.
This rg I-M situation was detennined to be a violation of Criterion XIII of /,
g Appendix B to 10 CFR 50 (Notice of Violation issued on May 31, 1983.
Reference:
Noge_e__of Violation 50-445183-18 and 50-446/83-12,
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The SRIC would note for the record that tfiere~is' ~
' little evidence that any items which indicated substantial deterioration from such storage conditions have in fact been installed in the nuclear power block.
It would appear that the various items involved have been cleaned and restored prior to installation such that they can perform the required function.
9.
Obsolete and/or Illegible Drawinos 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 we re inenmnlete in the title and ravition blneke ha C D T t' A 4 c e..e e a d.
8 the finding with supervisory personnel of the licensee's central document control center who indicated that they had located the drawings identified by the CAT inspectors along with many more that were obsolete in other areas.
It was stated that distribution system for engineering drawings had become faulted by the simple volume and by the need for so many points of distribution and audit verification thereof. Since problems are obviously still present, it was detemined that the licensee had violated Criterion VI of Appendix B to 10 CFR 50 (Notice of Violation was issued on May 31, 1983.
Reference:
- Notice of Violation 50-445/83-18 and 50-446/83-12, item 3) and that substantial steps would be required to correct the problems.
10.
A11ecations Relative To Improperly Suoported Items In The Control Room The president of CASE in a letter dated March 11, 1983, addressed to Mr. Richard C. DeYoung, Director of the NRC Office of Inspection and Enforce-ment, indicated that CASE had received infomation from an unidentified sourci to the effect that:
a.
There is field run conduit above the control room supported only by wi. re.
- y y
b.
There is drywall (or sheet rock) thgis supported by wire.
c.
There may be lights that are supp'orted by dire.
The SRIC has examined the suspended ceiling and the area above the sus-pended ceiling in the control room area and has examined the pertinent engineering drawings depicting both in relation to these alkgetions with the following findings:
There is a considerable amount of both safety-related y.d nonsafety a.
related conduit in the area above the suspended ceiling. 'fbe safety-related conduit is supported by Seismic Category I supports typical J of those used in other areas of the facility.
The nonsafety-related conduits are generally supported by simpler and less substantial sup-ports that are typical of those that the SRIC has observed in large open factories and are not designed to seismic standards.
In each case examined, the non-seismic support was structurally paralleled with a small stainless steel cable that would assume the full weight of the conduit were the nomal support to fail in a seismic event.
b.
The drywall materials were found to be part of the suspended ceiling above the central part of the control room and to fom a part of the sloping wall area below the control room observation room.
These dry-wall materials have been securely fastened to a metal frame work
.(metal batten) which in turn is supported by conventional and ncn-seismic straps and wires to the concrete primary building. The frame work is also attached to a system of stainless steel cables which in turn also attach to the primary structure such that if normal sup-ports fail during a seismic event, the weight of the framing and drywall will be assumed by the cabling thus preventing the materials from falling.
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The lighting fixtures in the control room are supported from an intermediate substructure of "unistrut" by light-weight conduit.
The substructure is likewise supported by the same type of conduit from the primary structure ceiling.
The conduit used appears to be the typical of that supporting the light fixtures in most offices with suspended ceilings.
Para 11ed with each conduit are two small stainless steel cables which would assume the load if the conduit or its attachment were to fail.
In the case of the actual light fixtures, the cable is attached to the light l
fixture at the edge of the reflector assembly.
i The SRIC would note for the record that above described design features appear to fully satisfy the intent of the licensee's commitment to comply with NRC Regulatory Guide 1.29, " Seismic Design Classification."
The licensee has used terminology in the classification system that is at variance with that of the regulatory guide but is explained and defined in Section 3.2 of the FSAR.
In essence, the licensee has defined all safety-related items that must remain fully functional during and after a seismic event as Seismic Category I.
Items not having a safety function but whose failure could damage components which have a safety function or cause injury to the occupants of the control room during an event are ref~ erred to as Seismic Category II.
In the case of the items involved in thi.s. allegation, all are Seismic Category II since their falling could
?
catise injury to the control operators.
The catMng system describe.d can be expected to prevent such a fall even though the normal supports rould possibly fail.
The stainless steel cable.used in this design feature,
)
which at a short distance away looks much like bright galvanized common steel wire, is of relatively high strength.
As an example, the test strength of an 1/8-inch cable is in excess of 1760 pounds. With four cables attached to a light fixture, two at each,end, the total support capability of the cables is over 7000 to use conventional. pounds. It is apparent that the designers ha've elected suspended ceiling and light fixture support techniques in order to use conventional and available materials and then provide a i
high strength backup support system i'n a seismic event.
No violations or deviations were identified during this special inspection l
effort.
11.
Placement ~ and Curing of Concrete During Freezing Weather During the limited public appearance portion of the Atomic Safety and Licensing Board (Board) hearing conducted on May 15, 1983, there were two references to the placing of concrete in freezing weather at the Comanche Peak Station which in turn lead to a question from the Board to the NRC i
staff as to whether there were any NRC personnel present with knowledge of the matter.
The two references are at 6106 and 6134 of the hearing transcript while the Board question is at 6109.
Also at 6109, an uni-dentified voice responded to the Board that the matter had been reported in IE inspection reports.
Research of the NRC inspection reports revealed that there had been such a discussion in NRC Inspection Report 50-445/77-01 which was categorized as an unresolved item pending the licensee's review and action on their finding of the problem.
The unresolved item was further discussed in NRC Inspection Report 50-445/77-04 with the closure of the item by an improvement'in the QA procedures.
t
'~
10 The SRIC has reviewed the matter, particularily with a view toward deter-mining whether the_ practices involved actually caused damage to the concrete involved.
The primary focus of NRC Inspection Report 50-445/77-01 (Details II, 4
paragragh 5) was directed toward two licensee " Site Surveillance Reports" which had been prepared approximately 2 weeks earlier than the inspection period covered by the inspection report.
The first of the licensee's reports (C-134-77) was directed specifically to findings by a licensee inspector that the surface temperature of Concrete Placement 101-2808-001 some 6 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 December 30, 1976, being completed at approximately 6:00 p.m.
Later, 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 B&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 idur.ing_the course gf this
' review to gain a clearer understanding of the' events which took place.
The licensee inspector stated during the interview that he was confident that his measurements were accurate and also stated that there was no phy-sical evidence that the concrete was frozen even though the surface temperatures were well below freezing.
Tne records also reflect that in order to resolve the issue, swiss hammer tests were run on the suspect areas after the concrete had fully cured.
These tests indicated that the suspect areas had attained strengths comparable to known properly cured areas, indicating that the concrete had not been damaged even though the possibility exists that it had been frozen for a period of time.
The records reflect that good concrete curing temperatures, i.e., above 400F i
were established and maintained shortly after the licensee's inspector's observation.
For the record, the SRIC would note that Placement 101-2801-001 took place in the Unit 1 Reactor Building.
The placement became the open area floor at the lowest full floor in the building.
This floor area, while suppor ting some equipment, serves primarily as a walk area.
As such, it is fully topped with an architural concrete making the structural concrete no longer accessable.
NRC Inspection Report 50-445/77-01 also discussed comparable events to that documented on Surveillance Report C-135-77.
One of these events was docu-mented by Surveillance Report C-068-76 on January 7,1976, and on B&R deficiency / disposition reports (now titled nonconformance reports).
These documents indicate that on January 7,1976, the surface temperature
'of Pl acement.105-2773-001, the foundation basemat for the Unit 1 Safeguards Building, were found frozen as evidenced by frozen wet burlap over certain areas that were not covered by insulating blankets.
The records also i
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11 reveal that the reported finding took place almost 7 days after the place-ment of the concrete.
Although the placement should not have been allowed to freeze in the time frame involved in accordance with the project speci-fication, the placement was accepted "use-as-is" on the premise that the curing temperatures during the 7 days were conducive to a good cure and that after 7 days there would be little free water in the concrete to freeze even though the burlap was froze. This conclusion is considered valid by the
.SRIC based on his review of publications of the American Concrete Institute and the Bureau of Reclamation.
Further, in responding to a separate finding that the field cure test cylinders made for the placement tested lower than allowed by the project specifications, swiss hammer tests were perfomed.
The swiss hamer tests indicated the concrete placement had full specified s trength.
Relative to the low reported strengths of the field cure cylin-ders, the SRIC would note that in his experience field cure cylinders will 4
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 DDR-C-460 which in turn discussed low tempera.tures during the curing per-Qid of three separate placements that were make-during the late-December 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 with good conditions until the cold weather occurred.
The NRC inspector involved in NRC Inspection Report 50-445/77-04 which closed the unresolved issue has state,d 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 i
concern was for prevention of repetition rather than any specific concern about the quality of the placements involved.
The SRIC would note for the record that there are no regulatory or industry prohibitions on placing concrete in cold weather conditions.
The American Concrete Institute and the Burepu 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 freezing provided that precautions are taken to prevent heat loss.
In mass concrete applications, the greatest danger to the concrete is on the exposed surface areas, particularily at corners and other edges of the placement.
It would be exceedingly rare for the mass of the concrete to freeze and sustain damage.
These publications also indicate that even if frozen, the concrete will nomally cure to full design strengths if temperatures con-p ducive to the hydration process are restored.
l 12.
Allegations Relative To The As-Built Verification and Design Verification Activities.
During April 1983, NRC personnel received allegations to the effect that i
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12 the QA group performing as-built verifications were not measuring support member dimensions and therefore, the " Vendor Certified Drawings" of the supports would not be accurate.
A second allegation from the same person indicated that the QA group charged with responsibility for verifying that design changes have been incorporated into the plant and that the inspection records for the installations accurately reflected that incorporation was being required with the use of a computer generated status document to
'make the verification of records.
The allegation was that the computer list-ing was faulty and therefore, the verification effort was equally faulted.
The SRIC has examined each of these allegations 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-i tion, the SRIC found that the allegation was and is factual.
The allegation.
however, does not appear to have any significant, impact on safety in that the as-built inspection was not developed to assure that the " Vendor Cer-tified Drawing" was an accurate representation of the support in all aspects.
The as-built program was established to assure only that the support loca-tion on the supported pip ~e and the direction of support is accurate for the purposes of perfonning the final pipe stress analysis.
The responsibil'-
ity for assuring that the support members and ofher characteristics of the individual support reflect the design drawing. requirements reside in other QA ~ roups associated with the fabrication and insta11ation' efforts" To also g
perform these functions in the as-built verification inspection would be a i
redundant inspection that would not contribute significantly to the safety function of any given support.
j l
Regarding the second allegation, the SRIC found that it too was factual but l
only at the specific time the allegation was made.
When making the allega-tion, the alleger provided the NRC personnel with a reference to a QC inspection 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 on a computer tabulation and that the report was being completed at the direction of the inspector's supervisor.
The original report was dated April 4,1983.
The permanent file copy was found to have been marked " voided" by the originating inspector as of May 20, 1983, with a notation that the report had been superceded by IR DCV-00423.
This latter inspection report was examined by the SRIC and found to document essentially the same inspection effort by the same inspector but without any notation of having been based upon a computer tabulation and without notation of apparent protest of directions given by supervision.
The SRIC interviewed the QC inspector who prepared and signed all of the reports noted above in order to ascertain what had and is transpiring in the QC design verification program effort.
The inspector stated that the attempt to use the computer based data in the performance of the assigned task was in error from the beginning because of errors by persons genera-ting the computer data. The interviewee stated that only the one verifica-tion effort had been done using the computer based data and that all prior an'd subsequent. verifications have been done by the assigned inspectors i
directly and personally examining the existent quality records in compli-ance with applicable QC procedures for the task.
He stated that the only
i 13 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 to give management a picture of the faults in the computerized data.
It thus appears that the design verification effort has been performed in accordance with procedures except for the one-time pertubation that was subsequent correctly reaccomplished in accordance with approved proce-dures.
No violation to NRC requirements were revealed during this special inspection effort.
l 13.
Improperly Certified Liquid Penetrant Examination Materials i
The CASE informed the Atomic Safety and Licensing Board by a letter dated May 18,1983, of a. potential problem with the liquid penetrant materials in
use at the Comanche Peak Station.
The letter stated that CASE had been made aware of the potential problem during a phone conversation with Charles A.
Atchison, who in turn learned of the " problem" from a Dallas area represen-
- ative of the Magna-Flux Corporation, the orginal manufacturer of the material.
The letter states that the problem surfaced Wy 7-to 10 days earlier.
Based I
on the date of the letter, it would seem that the problem arose between approximately May 8 to May 11,1983.
l 1
The situation bears close resemblance to the situation outlined beginning 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 no' ed t
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 l
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 " pen and ink" changes sometimes didn't match the type of material being certified.
A Notice of Violation was issued as part of the inspection report on the matter.
The licensee responded to the Notice of Violation by a letter dated December 21, 1982, wherein he stated that a supplier had altered the certificates but that the original manufacturer had been able to furnish valid certificates and further, that all future purchases would be direct from the manufacturer rather from a " middle-man" supplier.
The licensee also stated that specific receiving inspection pro-cedures had been implemented to prevent repetition.
NRC Inspection Report 50-445/83-10;50-446/83-05 documented verification that the licensee's actions were acceptable and the ~ matter was closed.
l It appears that the situation outlined in the CASE letter parallels the 1
NRC findings in all details except for the dates which probably arose j
.as a result of misunderstood or incomplete communications between the 6
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1G Magna-Flux representative and Mr. Atchison and/or with CASE, CASE also posed two questions on the matter as follows:
a.
Has an NCR been written on this problem?
Answer:
The above discussed inspection reports document a total of five NCR's that were issued.
b.
Has either TUGC0 or Texas Utilities or B&R notified the NRC of this j
problem?
Answer:
The roles of reportability were effectively reversed in that the NRC identified the problem and notified the licensee.
A need for further NRC action on this matter has not been identified and the matter is considered closed.
14 Penetration Seals
~
TAlisspecialinspectionwasundertakentoaschtainthevalidity.andsig-nificance of allegations received initially'by an NRC Headquarters Duty Officer on or about March 22, 1983, which were confinned and added to during i
a telephone interview with the alleger on March 23, 1983, by the SRIC and a NRC inspector assigned to NRC Region I.
The allegations, as understood by the SRIC, were:
I N
i 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 (i
flexible boots since the material supplier and BISCO are involved in a lawsuit.
The aggregate used in a radiation seal may separate giving rise to c.
improper personnel protection.
Since BISCO was and is on the Comanche Peak site installing seals, Region IV was selected for the purpose of this special inspection although the com-pany has involvement at several other nuclear power sites throughout the United States.
The SRIC obtained from the B_ISCO site manager all of the l
production and quality procedures applicable to the work'at CPSES as well as some that are not.
The alleger specifically mentioned that the NRC should review Procedures QC-507, SP-504, SP-505,.SP-505-1, and SP-505-2 in regard to the flexible boot overlap problem.
Each of the above procedures was in the books offered to the SRIC for review.
A brief discussion fol-1.ows as to the contents of these procedures:
a.
OCP-507:
This procedure covers the final inspection of installed l
. - ~
1 1 n r T T r " i
.i-s APPENDIX i
U. 5. NUCLEAR REGULATORY COMMISSION REGION IV NRC Inspection Report:
50-445/83-27
~
Docket:
50-445 Construction Permit: CPPR-126 Licensee:
Texas Utilities Generating Company (TUGCO) 2001 Bryan Tower Dallas, Texas 75201 Facility Name:
Comanche Peak, Unit 1 Inspection At:
Comanche Peak, Unit 1, Glen Rose, Texas Inspection Conducted:
y 10-Jul 1, and Sept,ber 9-22, 1983
/
ll l
m,
/f M"8& 85 Inspector:
2 e
F. C. StewaNtor. Insfett.or Date Reactor Project Section A Approved:
h M
.I O
3 D. M. Hunnicutt, Chief Date Reactor Project Section A
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r 2
O Insoection Summary Inspection Conducted May 10-July 1, and September 9-22, 1983 (Report 50-445/83-27)
Special, unannounced inspection of alleged improper construc-Areas Inspected:
tion practices expressed by Robert L. Messerly in an affidavit dated February 3, 1983, prepared for Citizens Association for Sound Energy (CASE) and in an interview conducted on April 14, 1983, by members of the NRC Office of Investigations Field Office, Region IV.,The inspection involved 120 inspector-hours onsite by one NRC inspector.
Additional information was received from an individual, who requested confidentiality, that a former B&R mi11 wright.had drilled holes through rebar without the required engineering approvals.
This su)plemental inspection involved 10 inspector-hours onsite by one NRC inspector.
Results:
Of the seven allegations regarding improper construction practices One expressed by Mr. Messerly, five were found to be unsubstantiated.
allegation regarding improper documentation was found to be substantiated, however, the error was p,roperly corrected by the licensee and appears to lack technical merit; and one allegation regarding the posting of NRC Form 3, O could neither be refuted nor substantiated, however, it too appears to lack technical merit.
No violations or deviations were identified.
Results of Supolemental Inspection The allegation that unauthorized cutting of rebar during installation of
" trolley tracks" in the fuel handling building is considered to be unsubstantiated.
No violations or deviations were identified.
9 0
1 3
I
- O Details b.
1
)
A.
Persons Contacted Texas Utilities Services Incorporated (TUSI) Employees l
- 8. G. Scott, Quality Engineering Supervisor G. Tanley, General Superintendent i
C. R. Hooton, Lead Civil Engineer,
R. M. Kissinger, Project Civil Engineer C. Fleming, Field Engineer Brown & Root (B&R) Employees W. Wright, Project Welding Engineer B.. Hauser, Field Engineering Superintendent C. Osborn, Tool Crib Foreman l
The NRC inspector also contacted other licensee and contractor employees l
i duri,ng the course of the inspection.
Prior to this inspection, separate and independent investigative interviews were conducted by members of the Of fice of Investigation Field NB:
dated May 20, 1983).
j Office, Region IV (see attached Report A4-83-005, Alleged Imoroper Construction Practices B.
The NRC inspector, through an interpretative review of Mr. R. L. Messerly's affidavit, dated February 3, 1983, and his statements during his' interview, l
determined that there were seven specifically alleged l
April 14, 1983, matters that required a detailed inspection effort to assess the me.rit and/or their potential impact on safety-related systems, component.
i l
l and structures.
d
.The seven areas of.NRC concern which Mr. Messerly alleged to have occurre are summarized as follows:
That B&R employees drilled undocumented and unauthorized holes that cut through reinforcing steel and that such drilling and cutting was 1.
Mr. Messerly provided a copy done at the direction of supervisors.
)\\Q of a personal diary which, he alleged, reflected undocumented and 7
unauthorized drilling.
That one of the main steam lines in Unit 1 was moved using the polar crane, thereby placing the section of pipe line in an unsafe stressed 2.-
condition.
O That he had cut through concrete reinforcing steel as directed by work instructions that were not in accordance with the approved 3.
method of documentation.
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That tubular hanger / support steel anchor bolt holes were enlarged
{. *,
l 4.
with a burning torch which he said was unauthorized.
i l
i That (Richmond) anchor bolts were not perpendicular to concrete f
5.
surface and, therefore, unacceptable.
I That stainless steel pipe attachments were welded on piping with'out e
6.
'i an inerting purge.
j That NRC Form 3, " Notice to Employees" was not posted on three main 7.
bulletin boards.
I C.
Insoection Findings Allegation 1 i
1.
Discussion Mr. Messerly stated that during his assignment as foreman over the first crew responsible for drilling through concrete and reinforcing steel (rebar) during installation of cable tray and pipe hanger j
l lk()
supports, he was ordered by his supervisors to loan out drill bits and/or. drill undocumented and unauthorized holes through rebar.
4 To further support his allegation, Mr. Messerly named B&R employees responsible for the alleged improprieties and those who could 4
substantiate his allegations. 1/
\\
i In addition, Mr. Messerly provided the NRC staff a copy of his personal daily diary in which he logged drilling of holes for j
electric cable trays / hanger supports and rebar cutting details. He stated that this diary also identified holes he drilled, in or j
through, rebar and concrete without having documentation and author-J ization.
)
2.
Chronological Findinos 1978-1982 In order to determine the magnitude of implication and the resulting i
l findings of Mr. Messerly's allegations.
i 1
l l
See attached " Assistance to Inspection Report," Report A4-83-005, dated 1/
~
May 20, 1983
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t The NRC inspector reconstructed, through the use of record archives and interviews with site personnel, the onsite construction activities and QA/QC program being implemented in the specific area of concern during the period 1978-1979.
f 3.
Rebar Cuttino Capabilities The NRC inspector found from B&R purchases that during 1975 through t
1982, the type of onsite equipment (drills) capable of cutting through rebar and available to craft personnel were restricted almost exclusively to the (water cool'ed) type diamond core drill bits (rebar eater) and associated drill motors, purchased from Drillco Equipment Company, Inc., (Drillco) Miami, Florica.
The Drillco water cooled diamond core drill bits purchased are hollos, hibular in shape, varying in sizes from 1/2" to 16" in diameter and from 2" tc 14" in length.
The drilling end has a series of carbide rectangular shaped j
teeth impregnathd with industrial diamond dust.
When worn, or dull, the bits can be reconditioned and reused.
The NRC inspector found that the initial core drilling requirements (1975 to 1978) were under the control of the concrete department.
Drilling was destricted' to investigative type core drilling (identif-(-
ing concrete honeycomb, voids or cold joints) in the base mats (NRC Inspection Report 445/446/76-04 dated April 20, 1976).
+
l
-Jn late 1977, record archives contain copies of the original " Core Drilling Procedure," MCP-13, dated September 27, 1977, and issued for implementation April 21, 1978.
The procedure was developed for core drilling through walls and slabs for the purpose of installing pipe Penetrations which fleeves, conduits, instrumentation sleeves, etc.
,were shown on drawings or included in design documents prior to
- concrete placement and inadvertently omitted, or penetrations which
. were added by the architect engineer (A/E) but for which the installa-
. tion information was not available to the field prior to concrete placement were covered by this procedure.
The procedure was applicable for'all core drilling required in the plant.
Core drilling was assigned to the mi11 wright department.
The procedure and its controlling document, " Core Drill Request Form," requires delineatic.i of exact location, size and rebar location, l
and contains review and approval signoffs.
This procedure continues to be the principal core drilling procedure (Revision 3, dated December 2, 1981).
Howev'er, current policy (as determined by the cognizant project civil engineer and reflected in documented records) i is the assignment of core drilling of 2-1/2" diameter and larger to e
the mi11 wright department and 1/2" to 2" diameter core drilling to the steel fabrication department drilling crew.
The NRC inspector
()
also noted that " Core Drilling Request Forms" do not imply rebar j
cutting; in fact, rebar cutting has for the most part, been avoided 1
where possible as stated by the project civil engineer during discus-sions with engineering personnel. This fact was observed by the NRC 1 J L. 'rT :r
I 1
- O inspector during his review of randomly selected " Core Drilling l
Request Forms" (1978 through 1982).
f, Construction records indicate that electrical cable tray, conduit hangers, and pipe hanger support insta11ations were initially started i
in late 1978.
This coincides with the formation of the steel fabrica-tion department pipe hanger crew (s), special drilling crew (headed up by Mr. Messerly), and the requisition of the water cooled diamond I
core drills and motors by the steel fabrication department (of which Mr. Messerly was a member) on September 6,1978.
A record search indicated a. Design Change / Design Deviation Authorization 2470, dated September 5, 1978, authorizing rebar cutting for Cable Tray Support j
No. 597.
This was an initial rebar cut made on September 9,1978, and l
identified by Mr. Messerly in his personal handwritten diary (see paragraph 6).
l The primary anchor and fasteners utilized at CPSES for the attachment of cable tray supports, conduit supports, pipe hanger supports, etc.,
'to concrete surfaces are the "Hilti" drilled-in concrete expansion anchor and " Richmond" screw anchor.
The Richmond screw anchor is positioned prior to concrete placement, whereas the Hilti requires i
concrete drilling and placement at the time of component installation i
(a licensee representative stated, that based on purchase orders, O
over one million Hilti bolts 1/2" to 1-1/4" in diameter, have been
~
[ \\0 t
installed to date).
Drilled-in expansion bolts are bolts havins
. expansion wedges so arranged that, when placed in a drilled hole and 3
the nut tightened, the wedges are expanded and the bolt is securely i
anchored.
4 The most predominant means of drilling holes into concrete for expansion bolts is the use of Hilti power drills, using Hilti carbide j
masonry bits of the same nominal size as the bolt.
This form of i
drilling does not have the capability to drill through rebar.
i j
In limited access areas where the Hilti power drills cannot be used, a flexible Drillco drive drill with drill press / vacuum base and Drillco water cooled carbide / diamond bits are used.
This form of i
drilling has the capability of drilling through rebar and was restricted I
to the steel fabrication department special drilling crew (headed by l
Mr. Messerly from September 1978 through October 1979).
For these two methods of drilling, no authorization is required for Hilti bolt installations (other than an approved hanger support l
installation " traveler" with its accompanying location drawings). A
' design change authorization is only required if relocation is beyond
,the drawing tolerance limits, or if rebar is encountered and requires cutting.
Construction quality programs of this nature rely heavily
' on each individuals personnal integrity to adhere to prescribed O
procedure requirements.
_, _ _ -,.iZlZ ~ Z7Z_. if~~~: r z.i -. C ; r _ _._. _ _.._._
9 7
- O i
A research of purchase orders for 1978 through 1979 conducted by the NRC inspector, indicated that only seven (Drillco) power drives that 4
facilitate water cooling capability were purchased during that time frame.
Two were issued to the mi11 wright department and five were issued to the steel fabrication department (under the control of Mr. Messerly).
Mr. Messerly requisitioned (from the B&R warehouse) three drill machines, with water cooling splash guards, and one flex shaft unit on September 6, 1978.
An additional flex shaf t unit was requisitioned by Mr. Messerly on October 6,1978.
In discussing the method of dr'illing with the Drillco water cooled diamond bits with cognizant site personnel, the NRC inspector was informed that when drilling with the diamond core bits, water cooling is mandatory.
The water provides two primary functions: it removes drilling debris (concrete / steel) as drilling progresses, otherwise the drill bit would bind; secondly and most important, without water cooling, the drill bit will readily " burn up," particularly when attempting to cut through rebar steel.
In addition, a drilling foreman stated that, drilling equipment is heavy and bulky and j
drilling set-up time (mounting to walls or ceiling) generally takes half an hour to one hour.
When drilling, the water cooling creates a concrete / water mist deluge requiring crew members (normally,two) to wear rain type outer protective clothing.
4.
Diamond Core Drill 81t Control i
In verifying the purchase and control of the diamond core drill bits, the NRC inspector reviewed 21 B&R purchase orders awarded to Drillco '
dating from January 13, 1978 through February 13, 1980.
The NRC inspector found that of the total 21 purchase orders, 10 I
requisitions were initiated by the steel fabrication department h
,j general superintendent, representing 293 core drill bit purchases, and 11 purchase orders were intiated by millwright supervisory personnel representing 122 core drill bit purchases.
In reviewing the accompanying warehouse requisitions contained in each of the purchase order files, the NRC inspector noted that in the case of the steel fabrication department orders, all requisitions i
bore the signatures of Mr. Messerly or his department personnel.
l Correspondingly all equipment ordered by the millwrights was issued j
to and signed for by a cognizant mi11 wright foreman.
The-NRC inspector conducted an inspection at each of the respective 3
j department tool crib areas (millwrights and steel fabrication).
The i
millwrights maintain a tool crib area enclosed by heavy gauge wire screen and a locked counter door access.
The tool crib attendant l
maintained a clip board type log specifically for the control of Drillco diamond core bits.
The log identified the individual, along with checkout and, return dates.
Entries in this log date back to
(
October 16, 1978.,
1
[
8 O
The steel f abrication department maintains a small separate building where the hanger installation crew controls the drilling equipment and bits.
The NRC inspector observed that the Drillco diamond core bits were separately stored in a large wooden cabinet with an accompany-ing combination lock.
The rtethod of control over drills and bits was discussed with the cognizart foreman.
The foreman stated that he had been in charge of diamond core bits and the fabrication department drilling crew since April of 1982.
He stated that he did not cut any rebar without an approved " request for rebar cutting" form, which he further demonstrated by utilizing an inprocess form dated June 14, 1983, No. 135.
The NRC inspector determined that this was in accordance with the prescribed procedure, CC-P-47, " Request for Rebar Cutting,"
dated June 17, 1981.
In interviewing former supervisors, foreme,n, and members of diamond core drilling crews 1/, all interviewees stated that the present method of controlling diamond bits has been in effect since the initial purchase of Drillco bits; i.e., only cognizant supervisors, foremen, or drill crew members have access to the diamond bits (those interviewed incl.uded five former members of Mr. Messerly's drill crew).
S.
Procedure Reviews and Procedure Imolementation During the inspection, the NRC inspector reviewed B&R procedures and procedural implementation applicable to concrete core drilling and drilling requirements for Hilti bolt installations.
Included in the review were the original versions of issued procedures 40 from archiye fiies snat were applicabie euring 1978 and 1979.
S Applicable procedures reviewed included the following:
B&R Procedure 35-1195-CEI-20, " Installation of 'Hilti' Drilled-In Bolts," dated May 31, 1978; B&R Procedure 35-1195-CEI-20, " Installation of 'Hilti' Drilled-in Bolts," Revision 8, dated January 26, 1983; TUSI Procedure QI-QP-11.3-2, " Cable Tray and Conduit Hanger Inspection," dated June 3,1978; B&R Procedure 35-1195-MCP-13, " Core Drilling," dated September 27, 1977; B&R Procedure 35-1195-MCP-13 " Core Drilling," Revision 1, dated April 21, 1978; TUSI Procedure CP-QP-11.2, " Surveillance and Inspection of Concrete Anchor Bolt Installation," dated December 13, 1979; e, - -,,,,
-e I
l 9
1 O-35-1195-CCP-47, " Request for Rebar Cutting," dated B&R Procedure l
June 17, 1981-TUSI Procedure QI-QP-11.2-1, " Concrete Anchor Bolt Installation, i
i l
dated December 13, 1979; and j
2323-55-30, " Structural Embedments."
G&H Specification The principal construction procedure applicable.for Hilti bolt 35-11 l
installation was B&R ProcedureSection 3.2.1 s'tates, " Expansion bolt holes shall h
t be drilled into concrete reinforcing steel unless approved by t e May 31, 1978.
l Thi: requi re-I, Gibbs & Hill, resident engineer or his representative."
i i to the ment has been retained in all subsequent (eight) rev s onsT l
i procedure.
Revision 8, dated January 26, 1983.
In discussing the method of " engineering approval" established in i
h NRC with the cognizant project civil engineer, t e d
inspector was informed that an " Interference Task Force" was e period 1978-1979 i
in September of 1978, composed of three TUSI project civil eng neers i
ith the cognizant who coordinated any design changes or rebar cutt ng wWhere interference between
?
l onsite, A/E Civil Design Engineer. expansion bolt and reinforcm i
. h was generally adjusted within the tolerances allowed by the des i
drawings, otherwise a design change / design deviation authorizat on (DC/DDA), design change authorization (DCA), or a component m The various forms'of ication change (CMC) was initiated and issued.
CA and design change documents have subsequently been reduced to the D CMC forms of design change approval.
ing steel cannot be avoided and the cutting of rebar is required, th l
approval authorization is initiated by the A/E site project civil The criteria i
engineer who evaluates all requests for cutting rebar.
for such evaluation is based on design parameters determined by the si-A/E. Final design approval for any rebar cutting remains the respon l
l bility of the A/E's New York office.
The A/E site project civil engineer maintains a CMC DCA
[(
The information on the DCA or CMC; for rebar cutting.
CMC 0188, dated October 3, 1978.
l i.e., number of rebar cut, size and location is transferred to a i
d separate set of building structural drawings especially establ she for showing "as-built" rebar cutting entitled "rebar drawings cutting In interviews with the cognizant A/E site project civil 1978-1979 1/, the NRC investigators were criteria."
informed that although requests to cut rebar came from a number o engineer assigned during different B&R craft personnel, he, almost always,.gave the approvin l
cac te "r r'v.
i c *i> cr eie 18 r c tti 9-further stated that he had no knowledge of I O l
I i
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10 The NRC inspector subsequently conducted ation E
O engineering approval. detailed review and documentation verif ca i
V. ' '.
y<
dures.
Messerly's Diary (Logl ly provided the NRC j
6.
14, 1983. Mr. Messerlog entitled, " Start l
i During the interview on Aprilinvestigators with a copy of his Detail." The diary j,
i tion entries on standard i
of New Crew and New Operation Rebar Cutt ng consists of 24 handwritten pag,es of column a1978, through October 8-1/2" x 11" paper dating from September 7,Five colu
/ hanger CA, or CMC number) were recorded b
support numbers); building location; re arIn add 1979.
iis position (floor, wall, flex, DC/DDA, D h
by Mr. Messerly.are interspersed thoughout the 24 pages.
inspector observed fV During a detailed review of the diary, the NRC) Mr, Messerly rec r
Of the i
that (barring errors due to legibil tya total of 2976 hol r/ supports.
2976 holes drilled, 280 rebars were cut.less than 10% of the holesidentified requiring cutting was encountered inAll rebar cuts, as noted l
O by either a OC/DDA, DCA, or a CMC.
drilled.
were identified, p
Twenty-one of these rebar cuts were related to non did not review these buildings; therefore, the NRC inspectorIn addition, of the 2976 holes l
as being in the particular authorizations. drilled, 247 were identified by Mr. Me (f
turbine building.
tting, the NRC 3
Of the remaining 63 documents authorizing re i
s for a diary.
The NRC g
comparative verification against Mr. Messerly s 32 authoriza-inspector verified 132 rebar cuts identified in the In addition, all 132 cut the specific building tions.
C were identified on the DCA or CM.
rebars were traced to, and identified on, structura iteria," with the corresponding authorizing document number.Messerly in his There was no rebar cutting, as identified by Mr.
ization number.
diary, that does not have a corresponding aut handwritten note'in the diary (assumed to be written by Adjacent to h ld) - floor S.W.I."t Number SW-2-0
. Ordered to drill by (name with eand Hanger /Suppor d "None 7".
the'date July 23, 1979, Under the rebar cutting column Mr. Messerly note,
p illed.
During d
Mr. Messerly also noted that eight holes were dr
- c
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~
11 lO an investigation of this particular support (SW-2-035-004-J03R t,r,..
)
g..,.
in the service water intake structure (S.W.I.), the NRC 30, 1980.
'f inspector found that the support was deleted on July D*
d The original bolt holes were subsequently grouted and conc surfaces painted.
h mark after his notation, Mr. Messerly was not a witness to t e actual drilling of the specific holes drilled by his crew members, and since seven persons formerly ass l
of unauthorized rebar cutting.'
4 this matter further.
j It was also observed by the NRC inspector that, during a d by the i
verification review of the 32 DCA's and CMC's identi service water tunnel alone. This was also mentioned byAll 48 rebar cuts were Mr. Messerly during his interview.
to the design change authorization documents.
i l
Although Mr. Messerly's diary consistently identified the
'~
percentage of rebar cut, the 2stablished G&H desig l
I O
bar.
The NRC inspector found no unauthorized rebar cutting identified by Mr. Messerly in his handwritten diary.
Concu1sion - Allegation 1 w
7.
j Mr. Messerly's allegation that B&R employees drilled undocumented a unauthorized holes that cut through reinforcing steel could not be 1
()
substantiated for the following reasons:
l g_;-
Mr. Messerly's statements lack sufficient specificity as to who he " loaned" the water cooled diamond drill bits to cut rebar, or I
a.
who specifically ordered him to cut rebar when and where.
l I
Former supervisors deny ordering Mr. Messerly to " loan" out drills or cut unauthori' zed rebar, nor did any of the five former l'
crew members support this contention.
In the event an unauthorized person did use a wa i
b.
be accomplished without the accompanying water cooling drive equipment, or if a drill bit was " loaned" for drilling concrete only, it is conceivable that drilling would be successful in defective without water cooling, but not necessari11y resulting 3
i l ()
workmanship.
4 t
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O al diary contained Although Mr. Messerly implied L.at his person t rebar cutting, F. <,
dentification of unauthorized and undocumenation, the NRC inspecto 0 ',7 c.
f unless shrouded by omission or misin orm uthorized by i
could not identify a rebar cut that was not a DC/DDA, DCA, or CMC.
bility/ control Although the method of diamond bit e-countaThe insp on individual personal 4
i exhibits a weakness, the need for rely ng
- .;,~-
d.
tting of rebar was integrity would not be diminished.
not, nor do not, suggest indiscri,inate cu idance of Documented records exhibit a purposeful avoFur C @T done.
4 strates that less than 10% of the recordeby h rebar interference.
I tting.
ified in this area of the There were no violations or deviations ident inspection.
A11ecation 2 4
Discussion 4 4# a4.
f ar se eriv et t e ia ai
<<4'a vit < r r rv 2. 1982.
1.
d the use of the that he had witnesseby a pipefitter O
interview on April 14, 1983, Unit 1 reactor containment building polar crane n a manner that put undue i
supervisor in relocating a main steam line m line 1/.,
tension on the pipe. persons involved with the movement of the Conclusion - A11ecation 2 t adicted his allegation 2.
Although B&R personnel named by Mr. Me iew of the onsite l
documented records regarding this matter.
fic 32-inch steam i
It was observed by the NRC inspector that th MS-1-RS-001-1302-2, b
1 and the' reactor building polar crane was ut The rmanent piping.
i to assist repositioning a section of th s pe cord of the licensee has maintained a documented engineering reThe NR large section of temporary specific line movement.
feedwater nozzle and previously i
the line was necessary in order that a l cate the permanent piping (attached to the steam generator used for water flushing) be removed and to re oThe reco d" due to the weight of l
section of the main steam line that had "sagge discussions with Westinghouse the temporarly installed flushing pipe.
i meeting notes (memorandum) which reflect (N i
s prior to the l
16, 1982 under the work activity, in addition to establ sThe line was moved on Januar l
l and acceptability.
i i
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$W supervision of the field mechanical engineering group, and was i
witnessed by an engineering representative who observed the installation and use of the dynamometer (to register crane lifting loads) throughout the operation.
The lift connections and applied forces were recorded and retained in the file.
The lifting points were consistent with~
the hanger locations to simulate the permanent support system.
The as-built configuration was analyzed for stress and the acceptability of the line confirmed.
In addition, the recent completion of the
" Reactor Hot Functional Test" did not reveal any undue stress conditions This allegation cannot be substantiated.
No violations or deviations were identified in this area of the inspection Allecation 3 1.
Discussion During Mr. Messerly's interview on April 14, 1983, Mr. Messerly (in referencing his personal diary) stated that he initially started drilling rebar based on the instructions of three part memos, DC/DDAS and subsequently the CMC.
the CMC was an improper document, he did imply that the DC/DD the three part memo were not the right documentation.
2.
Conclusion - Allecation 3 During the NRC inspector's review of Mr. Messerly's p holes (rebar cuts) he drilled on September 7 and 8,1978, for cable tray hangers 596, 642, and 643, Mr. Messerly made the notation "RFIC".
In researching the archive files, the NRC inspector found Request Nos. EH-14 and EH-15, dated Augustthe original Requ 29, 1978.
instructions authorizing rebar cutting contained in the RFIC wereAlthough the correct and authorized by the cognizant A/E design engineer, the RFIC document was not the " approved" method of authorizing a design change.
The NRC inspector noted that this documentation error was corrected by CMC No. 00766 issued on October 16, 1978.
The original document, the RFIC contained a note to this effect.
1978, Mr. Messerly's diary contains a reference to DC/DDA No. 2489On Septem for two rebar cuts for hanger No. 597.
In researching this particular DC/DDA, the inspector found that DC/DDA No. 2489 was not related to hanger No. 597.
identified the rebar cutting authorization.The NRC inspector found that DC The location and number of rebar cut was also traced to CMC No. 01146, dated September 20, 1978, and to the as-built building structrual drawings, "Rebar Drawings Cutting Criteria."
This allegation by Mr. Messerly was h
substantiated; however, the original documentation error was identified a short time after its occurrence and immediately corrected and did not impact on plant safety.
No violations or deviations were identified in this area of the i
inspection.
14
- O I
t A11eoation 4 I
i 1.
Discussion I
During Mr. Messerly's interview on April 14, 1983, and as stated in i-his February 3,1983 affidavit, Mr. Messerly indicated that anchor bolt holes in tubular steel hanger supports were enlarged with a i
burning torch in order to compensate for the angularity of the previously installed (Richmond) anchor bolts, rather than redrill the
- holes.
2.
Conclusion - Allegation 4 1
The results of the interviews of eight B&R employees, whose names were provided by Mr. Messerly and alleged.to have knowledge concern-ing the improper use of cutting torches on hanger material, is contained in the attached " Assistance to Inspection Report." 1/ Two individuals stated that they recall an instance during a redesign t
j modification of a hanger where it was discovered that holes had been enlarged by a burning torch, therefore, that portion of the hanger 1
was scrapped.
?
During the onsite followup inspection concerning this matter, the NRC
' ()
inspector discussed the use of cutting torches with the licensee's 4
welding engineers and fabrication department engineers.
The NRC inspector was informed that the use of cutting torches is not prohibited, provided it is done in accordance with prescribed B&R procedures and/or ASME,Section III, Subsection 4211 (thermal cutting).
In the l
case of tubular hanger installations, the preferred method of correction for hole misalignment is to drill offset hole (s).
This has been done 1
on many occasions via the design change. CMC document.
The cognizant l
project engineer, responsible'for approving and issuing CMC's for hanger modifications, stated that he knew of no CMC that involved authorization of hole enlargement or hole relocation on tubular hanger supports utilizing thermal cutting; however, thermal cutting has been permitted as necessary on other types of carbon steel supports, base plates, etc.
The NRC inspector conducted a walk-through of the containment building to examine accessible installed tubular hangers, specifically in the
, plant areas mentioned by Mr. Messerly duri,ng his interview.
The inspector examined approximately 60 hangers at the 905' and 860' elevations in the containment building.
Although limited in visual i
accessibility to each 1" or 1-1/4" drilled hole in each section of t
the tubular hangers, the NRC inspector did not find any hole that was enlarged by a cutting torch.
2 l
In addition, the NRC inspector discussed the subject of thermal l
cutting with the cognizant QC supervising inspector who was involved with inspections of tubular hanger installation during 1980-1982.
i, The QC supervi,sor stated..that neither he nor any inspector discovered i
~
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an installed tubular hanger hole having been enlarged by a cutting t
- torch, Based on the lack of specificity by Mr. Messerly, the lack of corroborative testimony by Messerly's witnesses, interviews by the NRC inspector with cognizant site personnel, and the (limited) examinations of installed hangers, this allegation could not be substantiated.
There were no violations or deviations identified in this area of the inspection.
Allegation 5 1.
Discussion During the interview on April' 14, 1983, Mr. Messerly stated that Richmond Insert anchor bolts installed between elevations 905' and 860' in the reactor containment building have not been installed perpendicular to the concrete surfaces and, therefore, are unaccept-able.
In addition, Mr. Messerly stated, "... whatever angle it is, we would drill it at that angle so that it would come through the tube (i.e., tubular steel) and when it comes out the other side of O.
the tube, it comes out as close to center as we could get it."
Mr. Messerly also stated, "Just go out there and pull any...
studded rod out of there, pull three of them and two of them is [ sic]
crooked."
2.
Conclusion - Allegation 5 During the NRC inspector's onsite follow up of this matter, the inspector found that the B&R Procedure CP-CPM 9.10. " Fabrication of ASME-Related Component Supports," (original issue 12/28/78) is the primary construction installation procedure to be implemented and followed by the hanger installation crews.
The " General Fabrication and Installation Requirements," Section 3.3.1.2 " Installation Tolerances," states in part,
" Field Fit Tolerances "The tolerances discussed above shall be maintained for support fabrication activities.
However, if during the installation, the support won't fit, the members may be " field fit" provided the piping and elevation tolerances shown below have been maintained.
All other tolerances regarding axial location, alignment, and base plate attachments must be adhered to unless
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otherwise noted on the drawing."
1 16 i O In addition, Section 3.3.2, states in part,
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. Surfaces of bolted parts in contact with the bolt or nut i
shall have a slope of no more than 1:20 with respect to a plane j
Where the surface of a high strength normal to the bolt axis.
bolted part has a slope or more than 1:20, a beveled washer l
l shall be used to compensate for the lack of parallelism."
l During discussions with the cognizant design engineers concerning the specific installation requirements relative to the limiting perpen-i dicular angle of the anchor bolts (Richmond Inserts), the NRC inspector was informed that the limiting perpendicular angle of anchor bolts (Richmond Inserts) to the concrete surface is, aside l
from the requirements of Section 3.3.2, is handled on a case-by-case No enlargement of the existing predrilled holes in the l
basis.
tubular steel is permitted without prior approval; however, numerous i
CMC's have been issued wherein offset holes have been authorize The approval is generally accompanied by the requirement that the l
large square bolt washer be welded in place using a 1/4" fillet on 2 1
The cognizant engineer further stated that the requirement above only applies to safety-related supports (ASME III, Subsection sides.
Enlargement of the NF, Classes 1, 2, and 3 component supports).
predrilled holes in the tubular steel for nonsafety supports is l
permitted without prior engineering approval.
4 Since Mr. Messerly specifically referred to the 860' and 905' elevations in the reactor containment building in his testimony, it was assumed by the NRC inspector that his specific concern was in l
reference to the permitted angularity of the safety-related Richmond Mr. Messerly was apparently of the opinion that l
Insert anchor bolts.
the anchor bolt should be precisely perpendicular to the concrete l
j surface, which appears to be a misunderstanding on his part of the Furthermore, Mr. Messerly's testimony installation specification.
reflected his awareness and knowledge of the procedural requirements, 1
therefore, it must be assumed that Mr. Messerly did not ignore
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procedural requirements and did not indiscriminate 1y enlarge pre-
- Further, drilled tubular steel holes in safety-related supports.
j that any offset or enlargement done by Mr. Messerly had prior engineer-As noted in Allegation 4, paragraph 2, the ing approval as required.
l NRC inspector conducted a limited visual examination of approximately 60 hanger supports at the 905' and 860' elevations in the containment During the examination, the NRC inspector found no hole building.
enlargements or anchor bolt angles (parallelism of bolt nut surf ace
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i to washer surface) that. appeared to violate the above installation It is concluded by the NRC inspector that this specifications.
specific allegation appears to be more of a design concern by Mr. Messerly, than an improper installation construction practice having been implemented by him.
17 4
- O The need for the Richmond Insert anchor bolt to be precisely perpendicular to the concrete surface is not required according to the documented criteria established by the licensee, therefore, this I
concern alleged by Mr. Messerly is not substantiated.
There were no violations or deviations identified in this area of the j
inspection.
i Allegation 6 L
1.
Discussion j
During Mr. Messerly's interview on April 14, 1983, Mr. Messerly stated, "There was a welding foreman out there that done [ sic) a lot of welding illegally without documentation, such as lugs on pipes I
without purges."
In addition, Mr. Messarly identified three l
individuals wNo would have knowledge of attachments (lugs) being i
welded on pipe without an inerting purge 1/, with specific reference to the 832' elevation in the reactor containment building.
l 2.
Conclusion - Allegation 6 As noted in attachment 1/, two individuals identified by Mr. Messerly were interviewed concerning their alleged knowledge of lugs improperly welded on to stainless steel pipe without purging the pipe when
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required.
Both interviewees denied any knowledge of impro'per welding i
activities.
During this inspection, the NRC inspector conducted an onsite follow up review of this matter.
The licensee's pipe welding procedures had been established prior to the initial piping installation early in the construction phase.
The procedures and implementation activities had been inspected and documented on numerous occasions throughout that phase of construction j
by the NRC senior resident inspector and independently by NRC regional staff personnel.
Therefore, during this inspection, the NRC inspector i
limited the review to pipe welding purge requirement established by i
the licensee.
The NRC inspector observed that t.he primary welding procedures associated with safety-related piping are B&R CPM-6.9, Appendix D.
" Welding and Related Processes," and B&R Inspection Procedure j
QI-QAP-11.1-26, "ASME Pipe Fabrication and Installation Inspection."
j Paragraph 3.5 of this procedure, states, in part, 1
" Purging shall be maintained for welding of attachments to stainless steel piping having a wall thickness of 1/4 inch or less for field welds only.
This may be waived on a case-byJcase basis by the PWE and Engineeririg.
This waiver shall be documented on the applicable WDC."
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la In discussing this matter with the cognizant project welding engineers, i
d the NRC inspector was informed that when a welding purge is requ re j
for attachment, welds, the requirement would be noted on the weld data card (WDC) and a " Hold Point" established for verification by a QC-However, in instances where the purge is waived, an The interoffice memo waiving the purge is attached to the WDC.
inspector.
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interoffice memo is controlled by a chronological numbering systemIt was furth i
' and filed within the permanent record files.
' out by the B&R welding engineers that the majority of stainless steel f
piping at the 832' elevation have pipe wall thickness in excess o l
be the limiting 1/4" wall, therefore, an inerting purge would not
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required for weld of attachment lags.
Based on the fact that prior NRC inspections have not 1 entified a d
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concern in this ' area, that Mr. Messerly's allegatio l
etc.), that the majority of stainless steel piping at the 832' elevation exceeds 1/4" wall thickness, and that perso l'
substantiated.
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There were no violations or deviations' identified in this' area of i
inspection.
>LO Allegation 7 I
1.
Discussion It was observed by the NRC inspector in Mr. Messerly's affidevit of 14, 1983, he February 3, ~1983,'and during his interview on Aprilstat y
3, " Notice to Employees," on three main onsite bulletin boards.
i Conclusion - A11ecation 7 2.
The Code of Federaf Regulations, Part 50 (10 CFR 50), was revised by The change 47 FR 30452 to add 10 CFR C0.7,, Employee Protection."and had an l
2 l
14, 1982, was published JulyAn important. element of the change is that of a require-ment to post NRC Form 3 at locations where the form 1982.
l work.
1 During a prior review of this matter by the NRC senior resident 50-445/83-03; in.$pector (SRI) (see NRC Inspection Reportthe NRC Form 3 was observed by dated March 28, 1983),
However, the precise.
50-a46/83-01, l
the SRI to be posted in early January,1983.
date (between October through January) of the posting of NRC
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14 Form 3 could not be established.
B&R personnel records indicate
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that Mr. Messerly was terminated on December 6, 1982.
The allegation cannot be refuted nor substantiated.
Furthermore, the matter lacks any technical merit relative to an impact on the safety of the plant.
There were no violations or deviations identified in this area of the inspection.
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20 SUPPLEMENTAL INSPECTION September 9 - 22, 1983 1.
Discussion As noted in the attached assistance to NRC inspection report,
" Supplemental," dated September 7,,1983 2/, during the course of an unrelated investigation, information was received that a former B&R mi11 wright had drilled holes through rebar without the required eger,ing authorization.-
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During the period September 9 - 22, 1983, the NRC inspector conducted an t
onsite follow up on this matter.
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From the information provided by the interviewees, the NRC inspector
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identified the specific " Trolley Tracks" 2/, as the drum and spent filter
,['t$ q handling equipment, liner transfer trolley process aisle rails, located
~~= % v,3 on the 810'-6" floor level, in room 252, of the fuel handling building.
Q :s 2S 5?dS The system is currently in the preoperational testing phase; howeveh,
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} this system is not a safety-related system.
In reviewing the jy ' g construction documentation records regardin'g the'insta11ation af the rail J
assemblies, the NRC inspector found that the rail base plates, rail clips, drilled Hilti anchor bolts, and rails were installed per drawing,
" Anchoring Details'for Radwaste Solidification Sys)em," Figure 39, Sheet 5 of 5, and by direction of Design Change Authorization (DCA) 7043, Revisions 4, 8, and 9, dated October-22, 1980, October 28, 1982, and November 11, 1982, respectively.. It was observed by the NP.C inspector that Drawing Figure 39, Sheet 5 of 5, container the following pertinent notes, "2:
Expansion bolts and base plate may oe moved in east-west direction to avoid interference with rebar running in north-south direction." and, "3:
For rebar running in, east-west direction, holes may be drilled through the uppermost #18 bar 9 only one rail location and expansion bolts shall be installed through the hole (it is assumed that bar interference shall occur at any one rail only)."
2/ See attached assistance to inspection report " Supplemental," dated September 7, 1983, Report No. A4-83-005.
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In addition, Revision 8 of DCA 7041 directed the addition of extending the length of the rails f rom the original 24'-3" long to 27'-6" (3'-3" section added to east end); also, Revision 9 permitted the modification of Hilti bolts (shortening) to, avoid cutting any additional rebar.
The NRC inspector met with the superintendent of the millwright department and interviewed millwright craft personnel that were directly involved in installation of the rail assemblies.
During the interviews, the NRC inspector found that the rail assemblies were installed during two different time periods.
Although actual dates were not established, it appears that the initial 24'-3" rail sections were installed in late 1982 and the 3'-3" extension sections were installed early in 1983.
The individual interviewed on September 1, 1983 2/, stated that he was not aware of the 3'-3" extension of the rails; therefore, his reference to his work activities involved only the installation of the initial 24'-3" rail sections.
In addition, it has been established that, aside from the core drilling foreman, five millwrights and one millwright foreman were directly involved in the installation of the base plates and rail assemblies.
~
(Three of the millwrights and the millwright foreman were individuals also interviewed.)
2.
Insoection Findinos As a result of the onsite followup inspection, records review, and interviews wit,h personnel, the inspection findings are as follows:
a.
As stated by the millwright interviewed on September 1, 1983 2/, and acknowledged by other millwrights, only the east-west, #18 rebar, running parallel with the east-west rail, was drilled through to accommodate the 1/2" Hilti bolts which secure the rail base plates to the 810'-6" floor.
This rebar cutting was authorized per Note 3, Drawing Figure 39, Sheet 5 of 5, DCA 7041.
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'O The alleger stated' that the 3'-3" extension rails were insta i lled through for.
accordance with the DCA 7041, and that rebar was dr b.
i department the south rail Hilti bolts by the steel fabricat ondrilling during installation of the 3'-3" rail extension.
24'-3" The millwright foreman stated that during installation of the illing crew rail base plates, the steel fabrication department d c.
by 8
i hts to assist the himself, therefore, he assigned one of the millwr g b required drilling crew foreman in drilling the holes in which re arH being cut.
be cut per the DCA was cut, d
During the inspection, two of the millwrights interviewed state illing Hilti bolt that north-south rebar was encountered during drho tting of d.
the particular rebar was not permitted by the Dwas m i
9 of DCA 7041.
d The NRC inspector had a TUGC0 licensee representative locate an lt.
The bolt was verify the modification of the specific Hilti bolocated a
'the t
millwright's contention that no unauthorized rebar was,cu.
O f
In discussing the use of the core drilling equipment with the cr there is supervisory personnel, the NRC inspector was informed that h
re drilling e.
no hard set policy as to who can or who cannot use t e co h drilling equipment as long as the equipment is used properly and t e being done is authorized and directed by craft foreman orAs wit ling supervisory personnel. wherein he stated that when the core dril the foreman did not show up, he (the millwright) completed drilling September 1, 1983 2/,
h uth rail remaining (approximately 10) 1/2" diameter holes for t e so base plate Hilti bolts.
The NRC inspector found no evidence to support the allegation llation of the unauthorized cutting of rebar was done during instafilter handling equipment f.
" Trolley Tracks" for the drum and spent 3.
Results i
The allegation that unauthorized cutting of rebar s
aisle rails is considered to be unsubstantiated.
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