ML20044E457
| ML20044E457 | |
| Person / Time | |
|---|---|
| Site: | Sequoyah  |
| Issue date: | 05/07/1993 |
| From: | Blake J, Crowley B NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML20044E448 | List: |
| References | |
| 50-327-93-12, 50-328-93-12, CAL, NUDOCS 9305250033 | |
| Download: ML20044E457 (21) | |
See also: IR 05000327/1993012
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UNITED STATES
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NUCLEAR REGULATORY COMMISSION
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REGION II
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101 MARIETTA STREET.N.W.
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ATLANTA. GEORGf A 30323
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Report Nos.:
50-327/93-12 and 50-328/93-12
Licensee: Tennessee Valley Authority
6N38A Lookout Place
1101 Market Street
Chattanooga, TN, 37402-2801
Docket Nos.: 50-327 and 50-328
License Nos.: DPR-77 and DPR-79
Facility Name:
Sequoyah I and 2
Inspection Conducted: April 5-9 and 19-23, 1993
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Inspectors:
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B.
,C owley
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Da'te $igned
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Approved by:
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jNat.e/Blake, Chief
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Date Signed
rials and Processes Section
Engineering Branch
Division of Reactor Safety
SUMMARY
Scope:
This routine, annourced inspection was conducted on site in the areas of
Inservice Inspection (ISI) (Unit 1) and status of corrective actions for the
Confirmation of Action Letter (CAL) dated March 4,1993, relative to the Unit
2 fxtraction Steam pipe break and the resulting overvoltage condition.
Results:
In the areas inspected, one violation, inadequate PT inspection of safety
injection system piping welds - paragraph 2.c.(1) was identified. Although
the one violation and the problems with PT inspection of the reactor pressure
vessel RPV nozzle to pipe welds. indicated soti,e weaknesses in the ISI program,
overall, the program appeared to be functioning well.
It appeared that all
responsibilities were defined in procedures and instructions, even though
responsibilities were somewhat fragmented and spread over several organiza-
tions (lack of well defined ownership). Nondestructive examinations (NDE)
were being performed by qualified personnel in a conscientious manner using
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9305250033 930518
ADDCK 05000327
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approved procedures. Supervision and Level III examiners were consistently
involved in day-to-day NDE activities.
Relative to the corrective actions for
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the Extraction Steam line failure, no problems were identified. The licensee
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appeared to be taking appropriate corrective actions to identify the piping
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damaged by Flow Accelerated Corrosion (FAC) and improve their overall FAC
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program.
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REPORT DETAILS
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Persons Contacted
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Licensee Employees
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- R. Beecken Plant Manager - Sequoyah
- R. Bentley, Technical Specialist NDE - Level III Examiner
- J. Bynum, Vice President, Nuclear Operations
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- L. Bryant, Maintenance Manager
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- M. Cooper, Site Licensing Manager
N. Dietrich, Erosion / Corrosion Project Manager
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- R. Fenech, Vice President, Sequoyah
- T. Flippo, Site Quality Manager
- I. Heatherly, Nuclear Engineering, Principle Mechanical Engineer
- W. lustice, Technical Support Manager
- D. Lundy, Technical Support Manager
- M. Medford, Vice President, Nuclear Assurance, Licensing & Fuels
- R. Proffitt, Compliance Licensing Engineer
F. Scalise, Assistant.ISI Supervisor
- M. Skarzinski, Technical Programs and Performance
- R. Thompson, Compliance Licensing Manager
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- P. Trudel, Engineering Manager
- H. Turnbow, Manager - Inspection Services
- G. Wade, ISI Supervisor
- J. Ward, Manager - Engineering and Modifications
Contractor Personnel
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W. McBrine, Senior Project Engineer, Altran Corporation
K. Rhodes, Team Leader, ABB Impell Corporation
R. Aleksick, Jr., Engineer, ABB Impell Corporation
Other licensee and contractor employees contacted during this _ inspection
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included engineers, QA/QC personnel, craft personnel, security force
members, technicians, and administrative personnel.
NRC Employees
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- P. Kellogg, Section Chief, RII, DRP
- S. Shaeffer, Resident Inspector
- W. Hollant Senior Resident Inspector
- S. Sparks, Project Engineer, RII, DRP
- Attended exit interview
- Attended interim exit on April 9, 1993
Acronyms and initialisms used throughout this report are listed in the
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last paragraph.
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2.
Inservice Inspection (ISI) (Unit 1)
The-inspector reviewed documents and records, and observed activities, as
indicated below, to determine whether ISI was being conducted in accor-
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dance with applicable procedures, regulatory requirements, and licensee
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commitments. The applicable code for ISI is the American Society of
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Mechanical Engineers Boiler and Pressure Vessel (ASME B&PV) Code,
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Section XI, 1977 Edition, Summer 1978 Addenda, except that the extent of
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examination for pipe welds, categories B-J and C-F, is the 1974 Edition,
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Summer 1975. NDE techniques are in accordance with the 1986 Edition of
the Code.
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For Unit 1, the first 10-Year interval ends September 15, 1994. There-
fore, this is the last outage of the first interval. The end date is
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based on the extended outage from August 1985 to November 1988.
A total of 18 relief requests (151-1 through ISI- 18) have been submit-
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ted. The NRC has taken action on all relief requests as documented in
letters and Safety Evaluation Reports (SERs) dated April 19, 1990;
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February 7, August 21, and October 21, 1991; August 31, 1992; and January
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6, 1993. All relief requests have been granted, granted with comments,
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withdrawn, or determined to not need approval.
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The licensee performs their own ISI NDE in accordance with TVA procedures
using TVA and contractor examiners. The Site Quality Organization (5Q0)
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with the aid of a level III examiner from the Corporate Inspection
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Services Organization (150) is responsible for implementation of the ISI
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program. The corporate ISI/IST Programs Organization is responsible for
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issuing and revising the ISI program and necessary drawings.
ISO is
responsible for issuing and revising NDE procedures and Scan Plans. The
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reactor pressure vessel (RPV) inspections are performed by Southwest
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Research Institute (SwRI) using their procedures and examiners under the
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direction of licensee NDE personnel.
For inspections other than the RPV,
contract NDE examiners are furnished by ABB AMDATA and VOLT Technical
Services.
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a.
ISI Program Review (73051)
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The inspector reviewed the documents listed below related to the ISI
program to determine whether relief requests had been approved by
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NRR, the services of an Authorized Nuclear Inservice Inspector
(ANII) had been procured and the inspector was involved in ISI
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activities, the plan had been approved by the licensee and to assure
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that procedures and plans had been established (written, reviewed,
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approved and issued) to control and accomplish the following appli-
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cable activities: program organization including identification of
commitments and regulatory requirements, preparing plans and
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schedules, and qualification, training, responsibilities, and duties
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of personnel responsible for ISI; repair and replacement program
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requirements; NDE personnel qualification requirements; and guidance
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for identifying and processing relief requests.
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NP Standard Std-3.1, Revision 2, Quality Assurance Program
NP Standard Std-3.4, Revision 1, Corrective Action
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NP Standard Std-6.10, ASME Section XI and Augmented Nondestructive
Examinations
Surveillance Instruction SI-Il4.1, Revision 21, ASME Section XI
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Inservice Inspection Program Unit 1
SQN _ Unit 1 Cycle 6 Inservice Inspection Scan Plan, Revision 1,
including Revision Log
QMI 576, Revision 0, Qualification Review for Contract Suppliers of
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Quality Control (QC)/ Nondestructive Examination (NDE) Inspectors
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QMP 102.4, Revision 6, Qualification and Certification, Requirements
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for Nuclear Power (NP) NDE cersonnel
QMP 110.5, Revision 6, Admiristration of Nondestructive Examination
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(NDE) Procedures
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QMP 101.5, Revision 4, Interface with Authorized Inspection Agency
for ASME Sections III and XI Activities
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QMP 102.20, Quality Control (QC) Nondestructive Examination (NDE)
Monitoring Program
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SSP 2.9, Revision 3, Records Management
SSP 3.1, Revision 2, Quality Assurance Program
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SSP 3.4, Revision 6, Corrective Action
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SSP 3.6, Revision 3, Problem Evaluation Reports
SSP 6.9, Re: vision 3, Repair / Replacement of ASME Section XI
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Components
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N-GP-1, Revision 5, Marking and Identification Procedure for CSSC
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Components
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N-GP-6, Revision 5, Preparation of NDE Data Sheets
N-GP-7, Revision 1, Verification of Components Support Settings in
Accordance With ASME Section XI
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N-GP-8, Revision 1, Weld Reference System
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N-GP-9, Revision 3, Approved Materials, All NDE Methods
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N-GP-14, Revision 0, ASME Section XI Component Distribution
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Instruction
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N-GP-15, Revision 0, ASME Section XI Code Classification Criteria
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N-GP-16, Revision 0,Section XI NDE Programs Drawing Preparation
Program
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N-GP-18, Revision 3, Ultrasonic Testing Supplements
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N-GP-21, Revision 1, Evaluation and Resolution of Ultrasonic Piping
Data
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b.
Review of Procedures (73052)
The inspector reviewed the following NDE procedures to determine
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whether these procedures were consistent with regulatory
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requirements and licensee commitments. The procedures were reviewed
in the areas of procedure approval, requirements for qualification
of NDE personnel, compilation of required records, and division of
responsibility between the licensee and contractor personnel.
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addition, the procedures were reviewed for technical adequacy and
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conformance with ASME, Sections V and XI, and other licensee
commitments / requirements.
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N-PT-9, Revision 6, Liquid Penetrant Examination of ASME and
ANSI Code Components and Welds
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N-MT-6, Revision 6, Magnetic Particle Examination of ASME and
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ANSI Code Components and Welds
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N-UT-18, Revision 14, Manual Ultrasonic Examination of Piping
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Welds and Vessels with Wall Thicknesses 2 Inches and Less
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N-VT-1, Revision 18, Preservice and Inservice Visual
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Examination Procedure
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c.
Observation of Work and Work Activities (73053)
The inspector observed work activities, reviewed NDE personnel
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qualification records, reviewed certification records of NDE.
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equipment / materials, and reviewed evidence of overview of NDE
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activities, as detailed below.
The inspector verified: availability
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of and compliance with approved NDE procedures, use of knowledgeable
NDE personnel, and use of NDE personnel qualified to the proper
level.
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(1)
Liquid Penetrant Examination (PT)
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The inspector observed the in-process PT examinations as
indicated below. The observations were compared with the
inspection attributes of the applicable procedure and the ASME
B&PV Code to verify the performance of acceptable examinations.
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Examinations Observed
Drawina
Welds / Component
CHM-2333-C-8
SIS-295
CHM-2333-C-8
SIS-296
CHM-2333-C-1
SIS-330
CHM-2336-C-4
RHRS-177
CHM-2333-C-3
SIS-007
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CHM-2333-C-3
SIS-021
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CHM-2333-C-3
SIS-008
CHM-2333-C-3
SIS-001
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During observation of PT inspection of welds SIS-295 and
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SIS-296, the inspector noted that the welds were accepted by
the NDE examiner, although both welds had areas of bleedout at
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the weld edges that could mask relevant indications. The welds
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had not been properly prepared for PT inspection. After
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questioning by the inspector, a TVA level III examiner visually
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examined the welds and re-PT inspected an area exhibiting the
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most severe bleedout and agreed that the weld preparation for
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the two welds was inadequate to obtain an adequate PT.
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Subsequently, the weld edges were re-prepped and re-PT
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inspected. A rejectable linear (5/8" long) indication was
found at the edge of weld SIS-296.
Problem Evaluation Report
(PER) SQPER930100 was issued to investigate this Problem.
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Acceptance of the inadequate PT inspection is in violation of
the following:
Paragraph T-650(c) of Article 6 of Section V of the ASME
Boiler and Pressure Vessel Code, the applicable Code,
requires that broad areas of pigmentation that could mask
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indications of discontinuities are unacceptable, and such
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areas shall be cleaned and reexamined.
Paragraph 5.3 of licensee procedure N-PT-9, Revision 6,
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the applicable PT procedure, requires, that .... no
surface irregularities shall exist that could mask
indications caused by unacceptable discontinuities.
This inadequate PT examination is considered to be in violation
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of 10 CFR 50.55a(g) and is identified as item 327/93-12-01,
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Inadequate PT Inspection of Safety Injection Pipe Welds.
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the end of the inspection, the licensee had taken aggressive
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action to determine the root cause of the violation, the scope
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of the violation and all necessary corrective actions.
Incident Investigation Report II-S-93-022 was being issued to
documeint investigation of the root cause of the violation and
corrective actions.
In addition to observation of the above in-process inspections,
the inspector observed preparations for PT inspection of the
eight (8) RPV nozzle to pipe welds. The inspector noted that
these inspection activities were not well organized. The
inspections were delayed a number of times.
It appeared most
of the delays were related to discussions with Health Physics
(HP) about radiation protection requirements (whether
respiratory protection was required). On the second day of the
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inspection, after two of the welds hao been inspected on the
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previous night shift, the inspector attempted to witness the
inspection of welds RC-01 and RC-25 (including weld butter
RC-01-SE and RC-25-SE). The ISI supervisor stated that
respiratory protection was not required. However, at the time
of dress-out, HP stated that, since other nozzle welds were
being prepped for inspection and insulation was being installed
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on the two welds inspected the night before, respiratory
protection would be required. The NDE examiners dressed out,
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including respirators, and entered the RPV cavity, only to find
that the location of the nozzle welds relative to the RPV studs
did not match the location shown on the drawing. Therefore,
the examiners had to leave the cavity without completing the
inspection, and resolve the drawing discrepancy. The inspector
pointed out to the licensee the weakness of possibly
compromising the quality of a critical inspection (such as the
RPV nozzle to pipe welds) by further hampering the examiners
capability with respirators, when the inspections could be
accomplished without respirators by a slight delay of other
work in the cavity while the inspections are in progress.
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Before all inspections were completed, the work sequence was
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re-arranged such that the examiners did not have to wear
respirators. The lack of organization in this inspection
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activity indicated a weakness in the inspection program.
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(2) Magnetic Particle (MT) Examination
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The inspector observed the in-process MT examination of weld
FDS-10 on drawing CHM-2339-C. The observation was compared
with the inspection attributes of the applicable procedure and
the ASME B&PV Code to verify the performance of an acceptable
examination.
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(3) Ultrasonic (UT) Examination
The inspector observed a portion of the in-process 45* UT
examination of' welds FDF-10A and 11 on drawing CHM-2339-C,
SH 1, and FDF-130A and 131 on drawing CHM-2339-C, SH 2.
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inspector also observed the calibration process for the 45* and
60* angle beam inspections for these welds. The observations
were compared with the inspection attributes of the applicable
procedure and the ASME B&PV Code to verify the performance of
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acceptable examinations. See NRC Report 50-327,328/93-15 for
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additional inspection activity for these welds.
(4) Visual (VT) Examination
The inspector observed the in-process VT examinations as
indicated below. The observations were compared with the
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inspection attributes of the applicable procedure and the ASME
B&PV Code to verify the performance of acceptable examinations.
Examinations Observed
Drawing
Support / Component
CHM-2438-C-1
1-MSH-303
CHN-2438-C-1
1-MSH-302
1-H4-203-1
1-FDH-203
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CHM-2675-B-1
RCP3CSABLT(01-08)
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(5)
Personnel Qualification / Certification
The inspector reviewed personnel qualification documentation as
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indicated below for examiners who performed the examinations
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detailed in paragraphs (1), (2), (3), and (4) above. These
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personnel qualifications were reviewed in the following areas:
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employer's name; person certified; activity qualified to
perform; current period of certification; signature of
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employer's designated representative; basis used for
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certification; and, annual visual acuity, color vision
examination, and periodic recertification.
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Examiner Records Reviewed
Method
level
Employer
Number
II
AMDATA
4
II
AMDATA
6
II
AMDATA
6
II
AMDATA
3
II
VOLT
4
11
VOLT
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(6)
Equipment Certification Records
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Equipment / material certification records, as listed below, for
materials used in the inspections detailed in paragraph (1)
above were reviewed to ensure compliance with all applicable
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requirements.
Eauipment Type
Eauipment Identification
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Penetrant Cleaner
Batch 91M0lP
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Penetrant Cleaner
Batch 89K03K
Penetrant
Batch 89K04K
Penetrant
Batch 87L010
Penetrant Developer
Batch 89K13S
UT Transducer
Serial 33548
UT Transducer
Serial 33596
UT Transducer
Serial 40651
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UT Transducer
Serial 40539
UT Instrument
Serial E17128
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UT Instrument
Serial E17018
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UT Couplant
Batch 093001
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UT Calibration Block
Serial SQ-61
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UT Calibration Block
Serial SQ-62
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(7) Overview of NDE Activities
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During observation of the above NDE activities, the inspector
evaluated the extent of overview of ISI NDE activities by
supervision and level III examiners.
In addition, Monitoring
Reports SQN-ISI-1,2,3,7,8,9,27,and 28 were reviewed. During
field observations, the inspector noted extensive Level III
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involvement and overview in day-to-day activities. The
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monitoring reports reviewed indicated detailed involvement of
supervision and Level III examiners in overview of NDE
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activities.
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RESULTS
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In the areas inspected, one violation, as detailed in paragraph c.(1),
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was identified. Although the one violation and the problems with PT
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inspection of the reactor pressure vessel RPV nozzle to pipe welds
indicated some weaknesses in the ISI program, overall, the program
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appeared to be functioning well.
It appeared that all responsibilities
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were defined in procedures and instructions, even though responsibilities
were somewhat fragmented and spread over several organizations (lack of
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well defined ownership). Nondestructive examinations (NDE) were being
performed by qualified personnel in a conscientious manner using approved
procedures. Supervision and Level III examiners were consistently
involved in day-to-day NDE activities.
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3.
Corrective Actions For Confirmation of Action Letter (CAL) Relative to
Unit 2 Extraction Steam Pipe Break
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On March 1, 1993, a 10-inch extraction steam line feeding the No. B2
feedwater heater ruptured resulting in steam being introduced into the
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main generator instrumentation cabinet causing a voltage excursion in the
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generator output and a manual operator trip of the reactor. The NRC
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performed a special Augmented Inspection Team (AIT) inspection of the
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event. The inspection activities and findings are documented in NRC
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report 50-327, 328/93-10. A CAL was issued on March 4, 1993. Most of
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the corrective actions specified in the CAL were in process and were
inspected during the AIT inspection. The purpose of the current
inspection was to review the status of the actions identified in the Cal
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and inspect activities relative to the status of piping subject to FAC,
including improvements in the FAC program.
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Confirmation of Action Letter
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The inspector reviewed the status of the actions identified in the
CAL and TVA referenced letter. The following is a summary of the
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actions stated in the letter and results of this review:
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Perform a complete review of the event, including a
determination of the piping failure mechanism.
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This action has been completed. The review is documented in
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TVA Incident Investigation Event Report (II) S-93-009. The
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piping failure mechanism is documented as FAC in TVA's
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Metallurgical Report. Review of this report is documented in
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the NRC AIT Report (93-10). The inspector reviewed TVA II
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Report S-93-009 and found it to contain a detailed review of
the event, including root cause assessment.
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Complete permanent repairs to piping that is currently under
temporary repair due to erosion / corrosion (E/C) effects.
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These repairs are still in process.
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Review your E/C program to identify any weaknesses, including a
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third party evaluation of the E/C program.
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The review to determine weaknesses is identified in TVA II
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Report S-93-009.
In general, the weaknesses identified
correspond to those identified in the NRC AIT Report. The
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third party review is being performed by the Electric Power
Research Institute (EPRI).
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Conduct field measurements of piping to assess the state of
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piping subject to E/C effects.
Field measurements are still in progress. See details in
paragraph c. below.
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Resolve all overvoltage issues which occurred as a result of
the transient, including conducting surveillances of selected
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solid state devices.
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The testirg to satisfy this requirement was essentially
complete, but the report had not been issued. No equipment
problems have been identified to date.
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Assess the impact of plant equipment in the vicinity of the
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ruptured pipe and assure there is no damage that could affect
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safe plant operation.
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This assessment is documented in TVA II Report S-93-008. The
report will be reviewed in a future inspection.
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Repair or replacement of all such piping and other hardware
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adversely impacted by the rupture or that were identified as
needing repair during the technical evaluations will be
completed.
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Pipe replacements are still in progress.
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The TVA letter referenced in the CAL indicated that insight
gained from evaluation of the weaknesses in the FAC program
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would be used to evaluate the effectiveness of other programs
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that may directly affect the safe operation of the units. Two
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action items related to this issue are identified in TVA II
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Report S-93-009.
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Action item 7 covers performance of an independent. review of
other site programs to determine if weaknesses similar to those
noted in the FAC program exist. This item has been completed
and a report, SQN - Technical Programs Review, dated March 31,
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1993, issued (see paragraph b. below for further inspection of
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this item).
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Action item 10 of the II report covers establishment of an
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organization for major programs that will ensure a program
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focus is maintained and ownership established. A Technical
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Programs and Performance Organization is under development.
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Staffing and position descriptions for the organization are in-
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the process of being developed.
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b.
Technical Programs Review
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As noted above, II Report S-93-009 indicates that independent
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reviews of other site programs would be performed to determine if
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weakness similar to those noted in the FAC program exist. The
inspector evaluated the review of the Section XI NDE program by: (1)
reviewing the TVA Report On The Assessment Of ASME/ Regulatory
Programs At Sequoyah Nuclear Plant (SQN), and (2) performing a
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detailed inspection of the ISI program (see paragraph 2. above).
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general, the ISI NDE program was fourd to be well defined and
functioning well. The following problems were identified with the
TVA assessment report and the Section XI Repair and Replacement
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Program
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The assessment did not appear to be a totally independent
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assessment. The team leader (Corporate ASME Section XI Program
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Manager) and one of the team members (Corporate ISI Program
Manager) have direct responsibilities in the ISI program. When
questioned by the inspector, the licensee stated that the
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involvement of these individuals was intentional because of the
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need to have personnel on the team with an overall
understanding of how the program functions. Although the
assessment was not totally independent, the licensee considered
their objective of assessing the adequacy of the programs was
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met.
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The inspector noted that one member of the team was a
contractor and he identified a number of problems with the
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Section XI Repair and Replacement (R/R) Program. A number of
his findings, although included as notes in the report, were
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evaluated by TVA and not identified in the report as problems
that needed corrective action.
Examples are as follows.
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Technical requirements of the R/R Plan are specified and
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reviewed by Planners, who are not sufficiently
knowledgeable in the technical requirements of Section XI
to be specifying and verifying implementation of technical
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requirements for R/R.
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Reconciliation of differences between
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fabrication / construction requirements for replacements and
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requirements used for original construction is apparently
!
,
not being performed for most replacements. The procedure
,
.
checklist contains one line item that is intended to
!
include reconciliation as part of the required suitability
evaluation.
It is not clear how this checklist line item
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'
function is performed, especially when done by a Planner
i
who has minimal experience in this area.
The contractor identified, in his notes, two problems with
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the 1992 steam generator feedwater elbow and transition
l
piece replacement packages, that were not identified in
j
the assessment report as problems needing corrective
actions. One pertained to a difference in impact
properties specified in TVA Design Criteria and DCN and
-
that specified in the Purchase Order and the Material
"
Certification. The other involved a difference in the
contract specified material specification and the actual
specification used.
,
.
4
..
-
.-
-.
.-
-
.
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,
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12
After questions by the inspector, the Assessment Team Leader
!
issued a letter to Engineering stating that these issues
required resolution.
!
!
The licensee had decided to bring their contractor back to the
site for further evaluation of his findings. He returned to
l
f
the site prior to the end of the inspection.
f
In addition to review of the Assessment Report, the inspector
i
discussed the R/R program with management and the R/R Coordinator,
,
reviewed the R/R procedure (SSP-6.9), and reviewed R/R documentation
i
for the 1992 replacement of steam generator feedwater elbows and
I
transition pieces.
Relative to procedure SSP-6.9, the inspector noted the following
!
,
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weaknesses.
3
!
(1) As identified by the contractor, many of the evaluation
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<
1
requirements in the procedure, such as suitability analysis,
failure determination, and root cause can be the responsibility
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1
of the Planner. Many of these functions should, more
!
appropriately, be performed by Engineering personnel. The R/R
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Coordinator indicated that in actual practice, he is actively
l
involved with planning to ensure that the activities are
i
accomplished correctly. However, this is not specified in the
'
procedure. The inspector pointed out that the procedure was
!
1
weak in details and definition of responsibilities in the areas
!
of failure analysis, suitability analysis, and root cause
,
analysis. There appears to be, based on procedure review, a
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1
lack of Engineering involvement in these activities.
!
1
.
(2) The procedure requires a suitability evaluation for
_
replacements per Appendix D.
Per Appendix D, the suitability
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l
evaluation is nothing more than comparing the replacement to
!
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the construction code, approved later code, or the original
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design requirements. This does not satisfy the Section XI,
4
i
Paragraph IWA-7220, requirements for a suitability analysis,
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,
As noted in paragraph 3.a. above, TVA has decided to establish a
!
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Technical Support Organization to manage technical programs. During
i
discussions with management of this new organization and managers
,
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now responsible for the R/R program, TVA indicated that the R/R
]
program will be evaluated and strengthened as required.
!
Relative to the 1992 replacement of the steam generator feedwater
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elbows and transition pieces, the inspector attempted to review the
,
,
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suitability analysis. The Appendix D Replacement Planning Checklist
i
for Unit I was reviewed, but could not be located for Unit 2.
]
However, as noted above, the Appendix 0 Replacement Checklist does
not satisfy Section XI requirements. At the conclusion of the
inspection, it was not clear whether Engineering reviews performed
for these parts, other than those documented on Appendix D, met the
!
,
)
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. _ .
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_
..
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13
suitability analysis required by Section XI. This matter and the
adequacy of the R/R program is considered unresolved pending further
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review to determine compliance with ASME Section XI requirements and
i
is identified as item number 327/93-12-02, 328/93-12-02, Adequacy of
Section XI Repair and Replacement Suitability Analysis.
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c.
Inspection of Piping for FAC Damage
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!
As noted in the NRC AIT inspcetion Report 50-327,328/93-10, in order
!
to determine the extent of vall thinning due to FAC, TVA developed a
!
screening, ultrasonic (UT) thickness measurement program. Details
i
and implementation of the program were inspected and are documented
in the AIT report. The inspector reviewed the results of these
>
inspections since the completion of the AIT inspection. The
!
following is a summary of the results through April 21, 1993:
INSPECTION POINTS
INSPECTIONS
INSPECTION POINTS
PROJECTED
COMPLETED
REJECTED
Small Bore
!
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Unit 1
515
262
20
Unit 2
576
576
103
'
large Bore
.
!
_
Unit 1
291
240
22
!
.
Unit 2
299
284
28
I
The " INSPECTION POINTS REJECTED" column is the number of inspection
i
areas that require repair or replacement because of being below
!
minimum or close to minimum wall.
Replacements using stainless
steel material for large bore piping and Chromium - Molybdenum
7
(CrMo) steel for small bore piping is in progress. The large
majority of significantly thinned piping is in the steam extraction
,
systems.
For Unit 2, the screening thickness measurements were
,
essentially complete except for a small scope expansion.
For the
'
small bore piping, the scope was still being expanded as new data
was evaluated.
For Unit 2 it was estimated that, for this outage,
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approximately 3000 feet of small diameter pipe will require
l
replacement. Replacements were estimated to be approximately 85%
complete. Approximately 28 areas in Unit 2 large bore pipe will
,
require replacement. This work was estimated to be approximately
l
40% complete.
Based on the inspection results, TVA had performed a wear rate
projection analysis and determined that for Unit 2, 39 areas in
small bore piping and 29 areas in large bore piping were acceptable
for a 6 month operating cycle, but not for a 24 month operating
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,
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_ . _
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'
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14
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cycle.
Similarly, for Unit 1, 90 small bore areas and 17 large bore
areas were acceptable for an 18 month operating cycle, but not for a
36 month period. This analysis was to determine if some pipe
-
replacements could be delayed for another cycle.
i
{
4
This inspection program and pipe replacement effort is on-going, and
'
therefore, the numbers above are changing daily.
In addition, as
i
new data is obtained and additional experience is gained, decisions
about whether to expend additional inspection time, or replace
!
sections of piping in one unit based on inspection results in the
'
other unit, or another train, also has an effect on the numbers.
-
i
In addition to the above review, the inspector observed inspection
activities on the Unit 2 turbine cross over/ cross under piping. The
i
primary inspection method for this piping is visual from the inside.
'
In general, the thinning in this piping was found to be minor.
t
However, some problem areas were identified at the pipe to tie bar
!
fillet welds at some expansion joints.
The pipe wall had been
eroded at the edge of the fillet welds.
In addition, approximately
20 linear indications (cracks and lack of fusion) had been
!
identified in vendor welds. These indications had not been
"
investigated to determine the depth and the full extent of the
,
problem. Westinghouse was contacted to aid in evaluation of these
-
problems and repairs were being planned.
j
The licensee provided the inspector Westinghouse Operation &
,
Maintenance Memo 034, dated April 27, 1983. This memorandum and the
'
forwarding letter dated June 13, 1983,
recommended that the turbine
4
cross under piping be inspected as a routine maintenance activity at
each refueling outage. Discussions with licensee personnel
-
indicated that this inspection had not been proceduralized and the
inspections were not being routinely performed. This failure to
!
perform recommended maintenance on the secondary plant equipment
appears to be a weakness in the maintenance program for the
l
secondary plant, consistent with weaknesses previously identified in
'
the erosion / corrosion program.
d.
Improvements to the FAC Program
!
1
(1) CHECMATE Activities
!
As noted in the NRC AIT inspection report, in parallel with the
i
above detailed screening inspections, ABB Impell is re-
'
performing the CHECMATE analysis, including re-modelling,
[
obtaining T , data, loading inspection data into CHEC-NDE, and
o
i
running the Pass 2 analysis. The inspector discussed these
activities with the Impell team leader and reviewed selected
a
'
,
data. The following summarizes the status of these activities
^
for Unit 2 (the Unit I analysis will be performed after the
l
Unit 2 analysis):
,
7
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. .
,
d
4
15
-
The modeling is being performed in accordance with the
requirements of the CHECMATE Manual and Impell experience.
-
The following systems are included in the current work:
High Pressure Extraction Steam to the #1,2, and 3 Heaters
and Reheaters
LP and HP Reheater Drains (Normal Operating)
MSR Shell Side Drains (Normal Operating)
Cascading and Forward Pumping Heater Drains (Normal
Operating)
Feedwater Pump Recirculation Lines to the Condenser
- 3 and #7 Heater Drain Pump Recirculation Lines
,
Steam Generator Blowdown System
Feedwater System
Condensate System
,
a
-
The modelling is being based on input from: drawings
(flow diagrams, isometric, physical arrangement, and
'
vendor), bills of material, the 100% power heat balance,
3
chemistry data, operations input, operating hours, and
walkdowns (where needed).
>
-
The modelling is being documented in design calculations
in accordance with TVA procedures and independently
'
verified by qualified Impell engineers and a TVA engineer.
In addition, the preliminary calculations for the first
five systems were sent to EPRI for review.
!
-
Based on the CHECMATE Pass I wear rate rankings and
j
industry experience, approximately 241 Unit 2 locations
i
,
j
were selected for gridding and thickness measurements for
!
determining T,,.
In general, a minimum of two components
{
4
were inspected in each CHECMATE analysis line. All
'
components downstream of normally operating control valves
j
were selected for inspection. Also, piping downstream of
i
each unique flow element, orifice or flow nozzle was
selected for inspection.
,
-
The measured wall thicknesses are fed back into the
CHECMATE model to perform a Pass 2 analysis to adjust the
i
Pass 1 CHECMATE wear predictions to provide plant specific
.
calibration of the model. The Pass 2 analysis was
essentially complete for all of the systems listed above
on April 22, 1993.
.
All inspection data is loaded into the CHEC-NDE program
and a review is made to determine which components appear
i
to be degrading from FAC. T , values are determined and
o
the erosion / corrosion analysis rerun to obtain new wear
,
4
..
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_ _
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.
._
__
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16
rates.
CHECMATE correlation plots are examined and
anomalous data points not within the EPRI acceptable
confidence band are dispositioned.
-
Program expansion is based on the EPRI guidelines of
measured wall thickness being less than 70% of the nominal
wall thickness. The following components (if not already
inspected) are selected for program expansion:
the
l
component downstream, all sister components in other
!
trains, and the next ranked component in that line based
i
on CHECMATE wear rate. Additional expansion locations are
selected based on the Pass 2 results. Components having a
remaining life of 18,000 hours0 days <br />0 hours <br />0 weeks <br />0 months <br /> or less to reach minimum
'
code wall thickness are reviewed to determine if
additional inspections are warranted. To date,
approximately 26 expansion locations have been identified
i
because of original inspection results from 13 locations.
i
-
During the AIT inspection, the team questioned the
licensee relative to thickness inspections and the use of
CHECMATE analysis for safety related piping inside the
,
containment. Although safety-related piping and main
!
steam piping inside the containment were modeled by
I
CHECMATE (original analysis by TVA), only one ASME class
'
component 1-1/2" blowdown line) on each unit inside the
containment had been inspected under the original program.
The reason safety-related piping and main steam piping
inside the containment was not chosen for inspection was
that sampling had been selected based on CHECMATE Pass 1
modeling, which only ranks components relative to
projected service life. Since the piping components
i
(feedwater and main steam) inside the containment have
'
long projected service life times, no thickness
.
measurement data was or would be collected to feed back
j
,
into the checkmate model to see if the projected service
!
life times are accurate.
i
!
-
When questioned by the AIT team, the licensee determined
the following relative to the 16" diameter feedwater
piping inside the containment:
The piping installed is schedule 80 (nominal wall of
,
0.844") with a calculated T,, of 0.694" including a
wear allowance of 0.080".
!
During the current wall thickness inspections of Unit
1 large-bore piping, the following feedwater piping
i
'
outside the containment was inspected: (1) bypass
piping - tees upstream and elbows downstream of
i
bypass valves, (2) feedwater regulation valve piping
- tees upstream of the valves, and (3) manifold
I
1
J
.
r-
.
,
17
arrangement to the feedwater heaters - upstream and
-
downstream of the FCVs, last tee and elbow of
,
manifold, and pump discharges. All of these areas
!
I
showed substantial margin when comparing the required
wall thickness to the measured wall thickness.
r
Also, during the AIT inspection, Impell ran a
CHECMATE Pass I analysis on typical feedwater system
t
piping components inside the containment using plant
r
history operating conditions. Although this data may
not be accurate until some actual measurement are
taken and input to calibrate the model, the analysis
indicated that the wear rate is very small (26 years
remaining to reach T.,).
During the AIT inspection,
the licensee stated that, for the new CHECMATE model,
some UT measurements of feedwater piping components
i
inside the containment will be made to input to and
validate the CHECMATE model.
,
During the current inspection, the inspector reviewed the
planned inspection locations, based on CHECMATE Pass 1
analysis, ident C ed by Impell for systsas inside the
containment. The following components had been selected:
Four 16" elbows at the inlet to
,
the steam generators
-
Four 16"X4" tees at the auxiliary
feedwater connections
One 18"X16" reducing elbow
Blowdown -
Three 2"X2" tees
Five 2" pipe locations downstream
of other components
.
'
One 2" elbow
t
In addition to review of the above process, the inspector
reviewed the new CHECMATE analysis (input and output data) for
the Unit 2 Feedwater System. The data reviewed included:
-
ECN 03 EROSION 22, Revision 0, Pass 1 Analysis
-
UT Thickness Results for Determination of T,,
-
Calc P2 EROSION 2, Revision 0, Pass 2 CHECMATE Model
!
-
Susceptibility Analysis
___
--
-_
-
,
.
3
i
18
The data was found to be detailed and well organized and
appeared to include the appropriate input data. The CHECMATE
analysis and the susceptibility analysis for plant piping
i
should be a valuable tool for use in predicting future FAC in
-l
plant systems.
>
(2) FAC Program
In addition to re-performance of the CHECMATE analysis, an
improved FAC program is planned. A program Manager, with
previous CHECMATE experience, has been appointed.
In addition
to interfacing with the current ABB Impell re-analysis, he is
in the process of formulating and writing procedures for an
l
enhanced program.
l
The licensee appeared to be taking appropriate corrective actions to
identify the piping damaged by Flow Accelerated Corrosion (FAC) and to
improve their overall FAC program. Weaknesses were identified in the
Section XI Repair and Replacement Program.
No Violations or deviations were identified.
4.
Exit Interview
The inspection scope and results were summarized on April 9, 1993, with
those persons indicated in paragraph I.
The inspector described the
areas inspected and discussed in detail the inspection findings listed
below. Proprietary information is not contained in this report.
Dissenting comments were not received from the licensee.
(0 pen) VIO 327/93-12-01, Inadequate PT Inspection of Safety Injection
System Piping Welds, Paragraph 2.c.(I)
(0 pen) UNR 327/93-12-02,328/93-12-02, Adequacy of Section XI
Repair / Replacement Suitability Analysis
5.
Augmented Inspection Team
ANII
Authorized Nuclear Inservice Inspector
i
American Society of Mechanical Engineers
Boiler and Pressure Vessel
Confirmation of Action Letter
DCN
Design Change Notice
E/C
Erosion / Corrosion
Electric Power Research Institute
Eddy Current Inspection
Flow Control Valve
High Pressure
Health Physics
II
Incident Investigation
e-
t
.:o-
l
,
'
.
?
19
!
Inservice Inspection
-l
'
Inspection Services Organization
Inservice Testing
i
Low Pressure
,
MBM
Manufacturing Burnishing Mark
!
Magnetic Particle Inspection
i
Nonconformance Report
!
i
NQA
Nuclear Quality Assurance
NRC
Nuclear Regulatory Commission
Office of Nuclear Reactor Regulation
!
PER
Problem Investigation Report
'
Preservice Inspection
Liquid Penetrant Inspection
j
Quality Assurance
!
QAI
Quality Assurance Instruction
-
I
Quality Control
QMI
Quality Management Instruction
QMP
Quality Management Procedure
RC
!
RII
NRC Region II
RHRS
Residual Heat' Removal System
i
l
Safety Evaluation Report
!
Surveillance Instruction
'
Safety Injection System
!
Sequoyah Nuclear Plant
SQ0
Site Quality Organization
l
Site Standard Practice
SwRI
Southwest Research Institute
i
Tennessee Valley Authority
-
Unresolved Item
,
Ultrasonic Inspection
Violation
i
Visual Inspection
1
WP
Workplan
l
l
i
I
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[
!
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