IR 05000424/1986078
| ML20214V277 | |
| Person / Time | |
|---|---|
| Site: | Vogtle  |
| Issue date: | 11/24/1986 |
| From: | Blake J, Sincule M, Vias S, York J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML20214V257 | List: |
| References | |
| 50-424-86-78, NUDOCS 8612090603 | |
| Download: ML20214V277 (23) | |
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- Licensee: Georgia Power Company
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Docket No.:
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b Inspections Conducteo: May 5-9, 198'6;7 June 2-6, 15-19, 22-25, 1986i+
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p July 7-11,20-24,)8-31,1986
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visionofReactorSafety, Region.II(
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This special, announced inspection.ss performed on Readiness,Msview
,g ' Module 16, Nuclear. Steam Supply System, in thec areaslof design progran e%duation ej and construction ppgram verification.
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Results:
Of the two areas inspected, one violation was found in each area (Violation, Ingrrect steam generator main ' steam nozzle load allowabbs and
refdence:specjfication, Paragraph 4a; Violat@' p \\ Failure to follow procedurc
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REPORT DETAILS 1.
Persons Contacted Licensea Employees
- D. O. Foster, Vice President and Project Support Managar
- D. S. Reed, General Manager of Quality Assurance
- E. D. Groover, Quality. Assurance Site Manager - Construction
- C. W. Hayes, Project Quality Assurance Manager
- R. W. McManus, Readiness Review (RR) Assistant Project Manager
- J.
L. Blocker, Project Support Staff
- G. A. McCarley, Project Compliance Coordinator Other Organizations
- W. C. Ramsey, Southern Company Services, RR Project Manager
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- H. Nixon, NISCO Project Manager
- J. G. Abbot, NISCO, QC Supervisor NRC Resident Inspector
- R. J. Schepens, NRC Resident Inspector
- Attended exit interview 2.
Exit Interview The inspection scope and findings were summarized on July 31, 1986 with those persons indicated by an asterisk in paragraph 1 above.
The licensee acknowledged the inspection findings listed below with no dissenting comments.
(0 pen) Violation 424/86-78-01, Incorrect steam generator main steam nozzle load allowables and reference specification, (Paragraph 4a).
(0 pen) Violation 424/86-78-02, Failure to follow procedure for documenting deviations, (Paragraph 4b).
The Licensee did identify as proprietary some of the materials provided to and reviewed by the inspestors during this inspection, however, details from those materials are not included in this report.
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Licensee Action on Previou~s Enforcement Matters This subject was not addressed in the inspection.
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4.
Program Verification (49051B, 49055B, 50071B, and 500758)
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This report represents the evaluation and direct inspection efforts of NRC l
Region II inspectors in regards to Readiness Review Team Module 16, Nuclear
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Steam Supply System, Section 6.0, Program Evaluation.
This section of the
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Module 16 Report described activities undertaken to ascertain whet 5er design
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and construction work processes were adequately controlled to ensure implementation of licensing commitments and conformance with project procedures and design requirements.
The section is divided into two subsections covering Design Program Verification and Construction Program Verification. The NRC Region II inspectors performed detailed evaluations of both subsections, which are described under the following two headings in this report.
a.
Design Program Verification (1) Review Introduction and Subsection Examination The Design Program Verification program was performed by the Readiness Review design verificatica team. The design verifica-tion concentrates on the design interface between Westinghouse, the nuclear steam supply system (NSSS) vendor, and Bechtel Power Corporation (BPC). The results of the review provided a basis on which to determine whether interface activities have been properly controlled.
Key areas were selected for review to ascertain whether the NSSS interface has been appropriately implemented and controlled. The design verification was performed in two phases by the Readiness-Review Team.
In Phase I, licensing commitments which are unique to VEGP, were reviewed to ascertain their implementation in project design documents.
In Phase II, design documents were reviewed to ascertain whether required design interface data has been properly transmitted, received, and implemented.
The NRC Region II inspectors concentrated on the Phase II part of the examination.
(2) Program Examination. In Phase II, the design interface between Westinghouse and BPC/GPC was reviewed. The Readiness Review Team selected eleven key areas for review to encompass the major NSSS interface activities.
The NRC Region II review included the following seven key areas:
o Piping stress analysis o
Reactor coolant loop equipment supports o
Accident analyses o
Instrumentation and control interface o
Electrical interface
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o NSSS equipment qualification and nozzle loads o
Equipment installation requirements.
The NRC Region II inspectors examined samples of design interface items in each of the key areas.
These samples included data provided to Westinghouse by BPC and data provided to BPC by Westinghouse. The review was directed ss much as passible toward changes in design which would result in interface activities. To accomplish this, the review also included samples of systems and components for which design changes had occurred that required additional design interface activities.
The Westinghouse /BPC/GPC interface activities were examined for proper and effective exchange of information between organiza-tions.
The areas considered were transmission of information required by other parties, receipt and correct internal distribu-tion of the information, implementation of information, and feedback, when needed.
The following sections discuss the review and results of the selected interface areas.
(a)
Piping Stress Analysis The Bechtel/ Westinghouse interface in the pipe stress analysis area was reviewed for proper control and effective exchange of data required by each organization and feedback of information. The review of the interface in this area concentrated on design changes which required interface activity.
The following areas selected for review encompass most of the interface activities in this area:
o Jet impingement loads on pressurizer surge '-
o RHR Recirculation and Hot Leg Injection o
Support load changes o
Support location change Design changes were selected for review and additional samples were included to ensure a review in each o# the above areas.
(1) Jet Impingement Loads -The following documents were reviewed:
o B-W 4358, dated August 3,
1984, this Bechtel document transmitted the complete set of jet impingement loads.
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o B-W 4080, dated August 9, 1984, this Bechtel document transmits pipe insulation data.
o W-B, GP 7483, dated August 30, 1983, surge line data accepted, except for some support stiffness.
o W-B, GP 8690, dated October 17, 1984, transmits-results of jet impingement evaluations o
W-8,GP 8621, dated September 13, 1984, suggests revised snubber size for unit 2 o
DWG 1X4DL4A17, R/4, Containment Building Piping Areas 4A, B, C, D, Level B - Class and Sections Reactor Coolant Loops The pressurizer surge line was addressed in each transmittal.
(2) RHR Recirculation and SIS Hot Leg Injection and Stress Analysis Input - The interface in the pipe stress analysis area was observed to be properly controlled and data effectively exchanged.
o B-W 4031, dated June 16, 1983, transmitting piping analysis Loop 4, RHR & SIS Hot Leg o
B-W 4042, dated June 28, 1983, transmitting piping analysis Loop 1, RHR & SIS Hot Leg Also transmitted were 3 isometrics for the above items:
1K4-1204-016-01 R/2 1K4-1204-049-02 R/1 1K4-1204-197-02 R/2 (3) Support Load Changes o
W-B, GP 9228, dated March 26, 1985, transmits analysis results, SIS Hot Leg Loops
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Attachment I-Support Design Loads and Displacements
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Attachment II-As analyzed support configuration o
Supports
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VI-1204-025-H001 R/1 - Rigid Strut and DCN-1
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VI-1204-025-H002 R/1 - Rigid Strut
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VI-1204-025-H002 R/4 - Snubber
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V1-1204-025-H012 R/1 - Snubber
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VI-1204-047-H003 R/3 - Snubber o.1d DCN-1
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(4) Support location change o
W-B, GP 8916, dated December 28, 1984, advises BPC of support load and configuration changes.
o ISO 1K4-1204-051-07 o
Henger V11204-051-H028 R/1 and R/2, DCNR-1 (b) Reactor Coolant Loop Equipment Supports. The review of the interface in the area of primary equipment support loads on the containment structure concentrated on the changes in the loads for the steam generator supports. Westinghouse again revised the support loads for this equipment in April 1985.
These new loads were transmitted to BPC.
The following documents were reviewed:
o W-B, GP-9325, dated April 23, 1985 transmitted Support Analysis WCAP-10734 Vol.2 o
BPC calculation X2CJ4.2.2 R/3 Steam Generator Support Embeds o
BPC calculation X2CJ4.2.2 R/4 revising Steam Generator Support Loads The interface in the area of primary support loads on the containment structure and the revision of loads were observed to be documented and controlled.
(c) Accident Analysis Westinghouse performs several types of accident analyses which require in formation from Bechtel.
-The containment pressure / temperature calculation data was selected for review.
The references from Westinghouse requesting data necessary for performing the pressure / temperature calculation are:
o Westinghouse letter GP-3602, dated January 25, 1980
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o Westinghouse letter GP-3623, dated January 31, 1980 A review was made of the data requested in the letters compared to the data supplied in letter designated as Bechtel log No. BW2999 dated August 15, 1980. The required informa-tion was supplied.
Additional information was supplied to Westinghouse by Bechtel after the VRR team review for Module 16.
The following two letters transmitted the additional or changed data:
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o Bechtel Letter Log No. 4647 dated July 8, 1985 o
Bechtel Letter Log No. 4655 dated July 31, 1985 (d) Instrumentation And Control Interface The interface in the instrument and control area was reviewed for incorporation of functional requirements specified by Westinghouse for Bechtel-designed systems, and for the input signals to be provided by Bechtel to the Westinghouse solid state protection system (SSPS), which is a part of the engineered safeguards features actuation system.
The following were reviewed:
Function Requirements
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The Westinghouse functional requirements which were to be implemented by Bechtel in the instrumentation and
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control area were reviewed using as an example the blackout signal for start of the turbine driven auxi-liary feedwater (AFW) pump. The following drawings were reviewed:
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o Drawing No.
1X4DB161-3 Rev. 14, P&I Diagram, Auxiliary Feedwater Pump System (Auxiliary Feed-water Pump Turbine Driven) System 1302 o
Drawing No. 1X5DN120-2 Rev. 4, Control Logic Diagram Auxiliary Feedwater System A review of these two drawings indicated that a blackout signal would initiate the start of the turbine driven i
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AFW pump and satisfies the requirement in Westinghouse document GAE/GBE-300/7 Rev. 3 i
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Solid State Protection System Input Signals an instru-mentation and control input signal for the main steam stop valve position required to be supplied by Bechtel to the Westinghouse Solid State Protection System were verified by reviewing the following drawings:
o Drawing No. 7243007 sheet 16, Logic Diagram (Westinghouse)
o Drawing No. 1X39-CD-B14A Rev.6, wiring Diagram Solid State Protection Cabling Block Diagram (Bechtel).
(e) Electrical Interface Samples of the interface in the electrical area were reviewed in three areas:
engineered safeguards features actuation system (ESFAS) valve and pump train assignments; Westinghouse electrical requirements for Bechtel-designed systems; and
. input signals to be provided by Bechtel to the Westinghouse SSPS which is a part of the ESFAS.
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Train Assignments Two pairs of valves and the two safety injection pumps were reviewed and verified as having the proper train assignment by reviewing the following drawings Valve or Pump No.
Drawing No.
Rev. No.
Valve 8801A 1X3D-BD-D02E
Valve 8801B 1X3D-BD-002F
Valve 8809A IX3D-BD-002V
Valve 88098 1X3D-BD-D02W
SI Pump 1 1X39-BD-001C
SI Pump 2 1X39-BD-D01D
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Power Requirements The AFW turbine driven pump flow path was verified as being independent from the diesel AC power source (a Westinghouse Steam Systems Design Manual requirement).
A review of the following three drawings verified that motor operated valve 1 HV-5106, which when opened allows steam to drive the turbine driven AFW pump, was dependent on 125 volt DC power:
o Drawing No.15DN120-2 Rev. 4, Control Logic Diagram Auxiliary Feedwater System.
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o Drawing No. 1X4DB161-3 Rev. 14, P & I Diagram Auxiliary Feedwater Pump System (Auxiliary Feedwater Pump Turbine Driven)
o Drawing No. 1X3D-BC-F02A Rev. 3, Elementary Diagram Auxiliary Feedwater System.
Also, the. independence of lube oil pump power and the lube oil cooling source from the diesel AC power source were verified by reviewing the following two drawings:
o Drawing No. C-6HMTA 321X88 (Bechtel Log No.
1X4AF03-130(1)-1)
o Drawing No. 103323E Sheet 1 of 2 from the Vendor Manual (Terry Corp.) Bechtel Log No 2X4AF03-229-2
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Solid State Protection System Input Signals and electrical input signal required to be supplied by Bechtel to the Westinghouse Solid State Protection System was verified by reviewing the following four Bechtel drawings:
o Drawing No. 1X3D-BD-801N l
o Drawing No. 1X3D-BD-B01P o
Drawing No. 1X3D-BD-801X l
o Drawing No. 1X3D-BD-801Y The reactor coolant pump under-frequency signal was shown as a separate channel on each of the drawings, showing that the signal had been implemented appro-priately.
(f) NSSS Equipment Qualification and Nozzle Loads Nuclear steam supply system equipment qualification was reviewed for nozzle loading on equipment from the attached piping.
The sample selected for review was the Steam Generator No. 1, Main Steam Nozzle loading.
For the steam generator, the allowable nozzle loading is specified by Westinghouse E-S Spe'cification No. 953236, Rev.
1, trans-mitted to Bechtel by GP-5121 dated 8/13/81. This document was referenced in Bechtel Log No. IX6AB12-32-1., received 8/25/81. The NRC Region II inspectors also reviewed Bechtel calculation X4CP-7110 R/2, page 10A/25A, which shows the steam generator nozzle loads and allowables. The reference given in the calculation was; "Give.: by W Table 2 of Specification X6AK03-46-3". This is tr. incorrect reference.
Also in the Fc Direction for the weight condition the allowable in the calculation was stated as (+/-) 15 Kips.
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'9 This is also incorrect, the proper allowable is (+/-) 10 Kips. The Readiness Review Team's Checklist, Item H-8 for the Steam Generator Main Steam Nozzle loads, was checked for this calculation and the Readiness Review Team did not identify the discrepancies.
Net identifying the correct allowable and the wrong reference of the Westinghouse Specification is a Violation of 10 CFR 50, Appendix B, Criterion III and is identified as Violation 424/86-78-01, Incorrect steam generator main steam nozzle load allowables and reference specification.
(g) Equipment Installation Requirements The interface in the area of equipment installation require-ments was reviewed in the control systems area.
Process Control Instrumentation was selected for review. The sample for review was selected to include various types of Class IE transmitters and to include installation.
The NRC Region II inspectors selected the RWST Level and pressurizer pressure transmitters for review.
It was during this portion of the review that the Readiness Review Team identified that the torque requirements for the transmitter bracket mounting bolts, shown on the Westinghouse drawing, were not included in the Bechtel drawing. Another example is the transmitter bracket mounting bolts shown on the Bechtel drawing as 1/4" bolts, not the 5/16" bolts shown on the Westinghouse drawing.
This was identified by the Readiness Review Team as Finding 16-13, which is discussed later in this report.
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The following documents were reviewed:
Bechtel Drawing No. Westinghouse Transmittal Letter 8765D52 R/3 GP-7029 8765D66 R/6 GP-7611 8765D67 R/5 GP-9316 8765D68 R/3 GP-2601 8765D69 R/6 GP-8327
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The installation requirements from Nuclear Instrument System (NIS) cable were selected for review. Westinghouse provides the installation requirements in the NIS Cable and Connectors Installation Control and Electrical System (C & ES) standard document.
This information was referred to in Bechtel Construction Specification X6AS01- *
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The following documents were reviewed:
o C & ES Standard R/5, transmitted by GP 3747, dated May 20, 1980 o
Drawing 1X3DF332 R/5 o
Bechtel Construction Specification X3AR01, Section E8, R/21 In the areas reviewed, except as noted above, the Westinghouse installation requirements were reflected in Bechtel design documents.
(3) Findings Subsection 6.1.4 of the Module 16 Report presents four findings disclosed by the Readiness Review Team in the Design Program Verification. Among these findings were one Level I, two Level II, and one Level III findings.
The NRC Region II inspectors performed a detailed evaluation on the one Level I finding and is described below:
o Finding 16-13--This finding noted that the installation bolting details, bolt size and torque values per Westinghouse supplied seismically and environmentally qualified electronic DP transmitters and pressure transmitters are not the same on the Bechtel drawings as those on the Westinghouse drawings.
The Readiness Review Team concluded that in the initial issue of the Instrument Installation drawings, certain vendor requirements for instrument installation were inadvertently omitted from review and were not incorporated.
The inspectors concluded that all changes would bo properly incorporated in the documentation under the program identi-fled for corrective action.
The examination of the Readiness Review Findings (Level I) in the Design Program verification disclosed no verification errors.
b.
Construction Program Verification (1) Review Introduction and Subsection Examination. The Construction Program Verification performed by the Readiness Review Team consisted of an evaluation of the installation of the nuclear steam supply system (NSSS) and related construction activities performed by Nuclear Installation Services Company (NISCO). The program had the objective of determining whether the construction control process functioned effectively and whether it insured acceptable installation of the NSSS components.
(2) P_rogram Examiration.
The assessments made and conclusions reached in Readiness Reviews verification program were divided into the two following major categories:
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- Hardware / Components o
. Programs / Procedures A further breakdown of the areas and attributes appraised is as follows:
(a) Hardware / Components The installation activities performed by NISCO for the.
following components were chosen:
- Reactor coolant pump No. 4 motor setting.
- Reactor coolant pump No. 4 support columns and tie rods.
- Steam generator No. 1.
- Pressurizer.
- Reactor pressure vessel (RPV).
- RPV head assembly.
- RPV internals.
- Bottom-mounted instrumentation:
A list of ibutes and activities examined for the 4.
installation of these components were:
- Documentation of activities
- Rigging and lifting
- Location and orientation
- Material / Component identification
- Clearances
- Welding control and inspection This is a composite list and not every component was examined for each of the attributes.
For example welding was not performed by NISCO on all of the listed components.
(b) Programs / Procedures Programmatic activities examined that supported NISCO's field installation activities were:
- Material Control,
- Document / records control,
- Personnel certifications,
- Nonconformance handling.
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A-itst of attributes examined under each of the programmatic activities were:
Materials Control
- Identification ~of the material (type, size, tag no.,
heat no);
- Appropriate signatures for inspections;
- CMTRs. if required, in conformance with specifications;
- Evidence of, acceptance or rejection.
Nondestructive Examination Test reports were reviewed for:
- Test part identification;
- Consumable materials;
- Governing code;
- Technique data;
- Examiner's qualification;
- Examination results;
- Legibility;
- Completeness.
Documents / Records Contrg
- Drawing numbers listed on the PCS were correct for the time of installation / inspection;
- Drawing revisions were documented on the PCSs;
- PCSs reflect appropriate information and entries;
- PCSs were available and retrievable.
Personnel Certifications
- Review of QC inspectors and NDE technicians qualifications to ascertain that QC personnel were qualified to the appropriate level for the QC test method being used.
Nonconformance Handling
- Appropriate disposition approval signatures;
- Proper closure and completion.
(3) Program Examination by Region II The Region II inspectors selected three NSSS hardware / components installed by NISCO.
These were installation of steam generator no.
1, pressurizer, and bottom mounted instrumentation guide
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While the Readiness Review Team divided the program l
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examination into two major categories, the Region II inspectors combined these two categories for the purpose of improving technical effectiveness.
(a)
Installation of Steam Generator No. 1 NISCO writes engineering specifications for installation of NSSS components. These specifications are then submitted to Bechtel and Westinghouse for approval. After the approval is received by NISCO then a working or installation document called a Process Control Sheet (PCS) is prepared. The PCS has step by step instructinns, QC sign off, ANI holdpoints, drawing numbers and revisions, receiving inspection report numbers, etc.
MISCO engineering specification ES-4028-3, Steam Generator Supports and Final Setting Procedure, was compared with the following process control sheets (PCS) to ascertain that all of the requirements stated in the specification were entered on the PCS's:
PCS No. 120-1 for Setting the Steam Generator PCS No. 958-13 for Steam Generator Vertical Column Supports PCS No. 958-13-1 for Steam Generator Vertical Column Supports PCS No. 958-25 for Steam Generator Upper Lateral Supports PCS No. 958-25-4 for Steam Generator Upper Lateral Supports PCS No. 958-25-6 for Steam Generator Upper Lateral Supports-PCS No. 958-25-7 for Steam Generator Upper Lateral Supports It was confirmed that the requirements were contained in the PCS's, that approved procedures were used, that approved hold points were established and observed, and that installation had been properly recorded and subsequently reviewed by NISCO.
The steam generator supports were supplied by Westinghouse and it was not necessary for NISCO to perform any welding for the erection of the supports.
A visual examination was performed by the Region II inspectors on the quality and sizes of some of the welds on the supports versus the drawing requirements.
The components of the supports inspected were identified as depicted on the drawings.
The supports for steam generator No. I were received on receiving inspection report (RIR) No. 5 and the documentation for the materials in this RIR were reviewed and accepted by NISCO. During the visual inspection of the supports, several components were noted for material acceptance verification.
The following materials were selected:
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Component and Westinghouse Serial Heat No.
Material Vertical Column S/N201-1 DWD A-588 DYF A-588 DKQ E-7018 S/N202-2 DWB A-572 Forging S/N6635 12-6935 A-471 S/N6623 22-5017 A-471 The certified materials test reports met the applicable material specifications. Also, materials A-588, A-572, and A-471 were code case No. 1644 materials which met the requirements of Regulatory Guide 1.85 Rey, 24 dated June 1986.
The foilowing deviation reports (DR) were reviewed:
OR No.
Subject NI-00003 Part not on approved vendor list.
NI-00066 Vertical column out of drawing tolerance.
MD-3559 Damage to steam generator scpports.
The deviations were ascertained to have the proper disposi-tion, completion, and closure.
(b)
Installation of Pressurizer Installation of the pressurizer involved a ring plate, and 24 bolts with nuts and washers.
In addition, NISCO installed the pressurizer seismic restraints that had been fabricated by Teledyne Brown.
NISCO engineering specification ES 4028-5, Pressurizer Final Setting Procedure, also contains information for erecting the seismic restraints and was compared with the following PCS's to ascertain that all of the requirements stated in the specification were entered on the PCS's:
PCS No. 130-1 for Setting the Pressurizer PCS No. 958-33 for Installing Seismic Restraints PCS No. 958-33-1 for Shims for Seismic Restraints PCS No. 958-33-2 for Locking Bars for Seismic Restraints PCS No. 958-33-3 for Arc Strikes on Seismic Restraints t
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It was confirmed thet the requirements were contained in the PCS's, that approved procedures were used, that approved hold points were established and observed, and that installation had been properly recorded and subsequently reviewed by NISCO.
No welding was performed by NISCO for the erection of the pressurizer' and seismic restraints.
However, a visual inspection was performed on a sample of the welds for the seismic restraints. The weld met the quality required by the specifications. The sizes specified on the drawing had to be altered in several places to allow the installation of the bolting. This condition and the dispositions were documented on a deviation report. An arc strike had been removed from lateral support (restraint) No. 801-1 and the area was dye penetrant inspected again during this inspection for evalua-tion.
The following DR's were reviewed:
DR No.
Subject NI-00008 Arc strikes on iateral supports NI 00030 Modification of welds, washers, and nuts to facilitate installation of lateral supports The deviations had the proper dispositions, completion, and closure.
(c)
Installation of Bottom Mounted Instrumentation Guide -Tubes and Supports The bottom mounted instrumentation guide tubes run from the reactor vessel to a seal table and are part of the reactor coolant pressure boundary.
The installation of the guide tubes and their supports involved ASME Code welding by NISCO.
The relevant ASME Code for the tubes is ASME Section III Subsection NB Class 1, and for the supports is ASME Section III Subsection NF Class 1.
The program examination on this hardware was first performed on the guide tubes and then on their supports.
Installation of Guide Tubes NISCO engineering specification ES-4028-12, Bottom Mounted Instrumentation Guide Tube Assembly, was compared with the following process control sheets (PCS) to ascertain that all of the requirements stated in the specification were entered on the PCS',
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i PCS No. 90-1 for Guide Tube Installation l
PCS No. 90-1-6 for Reworking Guide Tube Thimbles
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PCS No. 90-1-31 for NDE of Seal Table Welds It was confirmed that the requirements were contained in the PCS's; that approved procedures were used; that approved hold points were established and observed; and that installation had been properly recorded and subsequently reviewed by NISCO.
The following documentation for weld material was compared with the ASME Code requirements:
Weld No.
Filler Metal Type Heat No.
RIR 1A9 ER Ni Cr-3 A5006N382 113 1A9 ER Ni Cr-3 C4972G382 113 2F1 ER 308 L 26245
2F1 ER 308 L 05394
The filler metal met the ASME Code requirements.
The following documentation for other material in the guide tubes was compared with the ASME Code requirements:
Componenta Material Alloy Heat No.
RIR No.
Seal Table SA240 304 817925 172 Guide Coupling SA403 304 56873
Guide Tube No. 55N14 SA213 304L 465013
Guide Tube No. IJ8 SA213 304L 463582
The material met the ASME Code requirements There were 58 guide tubes and NISCO performed six welds on each tube from the reactor vessel penetration to and including the seal table weld.
The following NISCO welding procedures used for performing these welds were reviewed:
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WPS 438.2.2, Welding Procedure Specification for Manual Gas Tungsten Arc Welding (GTAW) of Inconel Bottom Head Adapters to Stainless Steel Instrumentation Tubing (Reactor Bottom Mounted Instrumentation Connections).
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GWP-BMI-1, General Welding Procedure-Reactor Bottom
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Mounted Instrumentation Connections.
WPS 80.2.5, Welding Procedure Specifications for Manual
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Gas Tungsten Arc Welding (GTAW) of Austenitic Stainless Steel Couplings and Seal Plate to Austenitic Stainless Steel. Instrument Tubing (Reactor Bottom Space Mounted Instrumentation Connections).
Engineering Specification ES56, Welding Filler Metal
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Control Procedure.
The PCS's were examined for determining if visual, fit up and dye penetrant inspections had been performed for the following welds:
Tube No.
Wold Nos.
Welder 8-6 1 thru 6 N-15 C-7 1 thru 6 N-12 D-10 1 thru 6 N-20 E-5 1 thru 6 N-15 F-1 1 thru 6 N-19 The PCS's indicated that the quality control inspections had been performed. Welder qualifications and ASME Code welding continuity records verified that the welders were qualified to make these welds.
The PCS's were also examined for x-ray (RT) inspections performed on weld No. 1 (reactor vessel penetration to instrument guide tubes).
The following sample was taken:
Accepted Tube No.
RT Report No.
B-6
D-10 125 F-1 126 In.the area of NDE the following NISCO procedures were reviewed:
Engineering Specification E.S. 100-5, Visual Inspection
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of Welds Engineering Specification E.S. 100-2, Liquid Penetrant
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Examination
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During a walkdown of the bottom mounted instrumentation guide tubes, a random sample of the welds were visually inspected and dye penetrant inspected again. Weld No.1 for 30 of the 48 guide tubes was visually inspected by the Region II inspectors with acceptable results.
However, a mechanical impression was found on tube No.48-P-4 approximately one foot below weld No. 1.
Evaluation of this potential tube damage was not documented by NISCO. The NISCO Engineering Specifi-cation E.S.142, Deviations and Corrective Action, Revision D, paragraph 4.1 defines a deviation as a deficiency in characteristic, documentation, or procedure which renders the quality of an item unacceptable or indeterminate.
Several tests had to be performed before acceptability of the tubing could be evaluated, e.g.
wall thickness, dye penetrant inspection, and checking to see if a metal ball of a specific size would pass thru tubing.
In addition NISCO needed to contact Westinghouse to determine if this condition was acceptable.
Therefore the potential tube damage was an indeterminate condition until all of the tests were performed and NISCO should have generated a deviation report according to their procedure E.S.
142.
Not following procedure and generating a deviation report is a violation of 10 CFR 50, Appendix B, Criterion V and is identified as Violation 424/86-78-02, Failure to follow procedure for documenting deviations. A second example can be found in a following paragraph.
Visual examination was also performed on a sample of seal table welds during the observance of dye penetrant inspection by a NISCO QC inspector.
Dye penetrant inspections requested and observed by the Region II inspectors are as follows:
Tube No.
Weld No.
46-J-1
30-F-3
28-C-7
35-D-8
4-H-6
57-H-6
Weld Nos. J10, 86, and H3 on the seal table were also dye penetrant inspected.
The tests were performed and evaluated to the NISCO engineering specification and ASME Code require-ment.
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The NDE certifications for the inspector who performed the test, and for two other inspectors found on the PCS's were examined and found to be qualified to the appropriate levels for the test method being used.
The following DR's were reviewed:
DR No.
Subject NI-00006 Lead-Copper go no go gage used on tubing NI-00014 Tube end prep damaged NI-00016 Linear indications on coupling NI-00085 Seal plate hole too small NI-00104 Wrong tube to seal plate hole The deviations had the proper disposition, completion, and closure.
Installation of Guide Tube Supports NISCO engineering specification ES-4028-13, Installation of Bottom Mounted Instrumentation Guide Tube Supports, was compared with the following PCS's to ascertain that the
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requirements stated in the specification were addressed on the PCS's:
PCS No. 90-2 for Support 1 (Group A) installation PCS No. 90-3 for Support 2 (Group B) installation PCS No. 90-4 for Support.3 (Group C) installation PCS.No. 90-6 for Support 5 (Group E1) installation It was confirmed that the requirements were contained in the PCS'S, that approved procedures were used, that approved hold points were established and observed, and that installation had been properly recorded and subsequent reviewed by NISCO.
The following documentation for weld material and welders who performed welding on the supports was examined to determine if ASME Code requirements were met:
i Weld No.
Filler Metal Heat or Lot No.
Welder BM181 E-7018 422X8361 N12 BM18144 E-70S2 97401 N26 BM18507 E-70S2 97401 N26 and N36 BM18876 E-70S2 97401 N26 BM181154 E-70S2 97401 N26 The weld filler metal met ASME Code requirements, and the welder qualifications and continuity records verified that the welders were qualified to make these weldr.
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The following documentation for other material used in the supports was compared with ASME Code requirements:
Material Hea t No.
RIR No.
SA-36 422P8311
A-588 411N0341 102 SA-307 KC3682 140 SA-36 422T5131 154 The material met the ASME Code requirements.
The following NISCO weld procedure used for making the support welds were reviewed:
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WPS 10.1.6, Welding Procedure Specification for Manual Shielded Metal Arc Welding (SMAW) of Carbon Steel to Itself on plate, pipe, tube, etc., 3/16" thru 5/8" thickness.
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WPS 10.3.1, Welding Procedure Specification for Gas Tugsten Arc Welding (GTAW) for Carbon Steel to Itself on Plate, Pipe, Tube etc., 1/16" to 5/8" thickness.
A walkdown of some of the supports for the BMI tube guides was performed in order to visually inspect some of the welds and some of the dimensional attributes for the supports.
Support 1 (Group A) the first support starting from the reactor vessel and going towards the seal table had 1153
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welds. A visual and dimensional inspection (fillet size) was performed on a random sample of 43 welds. Visual inspections were performed on three other supports in this system with the only apparent deviation noted on the support E-2.
Two
"U" bolts attached to the support structure steel for the
purpose of restraining the guide tubes appeared to have larger clear. aces between the "U" bolt and guide tubes. Tube No. 49-0-14 had greater than 1/8" clearance and tube No.
42-D-3 had greater than 1/4" clearance. Westinghouse drawing No. 1599E75 sheet 2 (Bechtel Log No. IX6AB08-34-8) showed 1/16" each side or total of 1/8" allowed clearance.
This deviation from the drawing had not been documented by NISCO.
This was identified to the licensee as the second of two examples of Violation 424/86-78-02, Failure to follow procedure for documenting deviations.
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The following D. R.'s were reviewed:
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DR No.
Subject
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NI-00070 linear indications where angle clips removed NI-00057 Missing hardware from BMI support steel NI-00107 Carbon content exceed chemistry requirement NI-00009
"U" bolts for supports undersize.
The deviations had the proper disposition, completion and closure.
(d) Findings Subsection 6.2 of the Module 16 Report presents five findings disclosed by the Readiness Review Team in the Construction Program Verification.
Four of the findings were in the d
paperwork category and one finding was in the programmatic category. There were no findings in the hardware category.
In the level of importance for the findings, there were no level I,
there were three level II and two level III findings.
The NRC Region II inspectors performed an evaluation on all five findings, but performed a more detailed review of the following two findings:
o Finding 16-6 This finding noted that the assembly sequence for the reactor pressure vessel internals was not in accordance with specification, and that the flange leveling requirements had not been implemented.
A review of the response indicated that the assembly activities would not affect final internal assembly and installation.
The upper internals assembly flange levelness had not been checked at the time of the Module 16 inspection.
A later check found the levelness to be within tolerance. A Deviation Report for the two items was dispositioned Use-As-Is.
o Finding 16-10- This finding noted four minor documenta-tion discrepancies with the radiographic reports.
None of these discrepancies questioned the quality of the weld. A fifth item of the finding involved five films of a similar joint configuration that displayed linear indication not noted on the radiographic reports. The linear indications were not addressed because they have been interpreted to be slag pushed up under the clip angle and out of the weld area, which is common with this joint configuration.
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Other ASME Code film was re-reviewed to ensure that linear indications had been addressed and the x-ray film interpretation was found to be in full compliance with Code requirement.
The examination of the Readiness Review Findings in the Construc-tion Program Verification disclosed no verification errors.
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