IR 05000395/1993011
| ML20035F197 | |
| Person / Time | |
|---|---|
| Site: | Summer  |
| Issue date: | 04/15/1993 |
| From: | Blake J, Economos N NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML20035F191 | List: |
| References | |
| 50-395-93-11, NUDOCS 9304210019 | |
| Download: ML20035F197 (12) | |
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- 4 NUCLEAR REGULATORY COMMisslOM l
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101 M ARIETT A STRE ET. N.W.
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Report No.: 50-395/93-11 Licensee: South Carolina Electric & Gas Company Columbia, SC 29218 Docket No.: 50-395 License No.: NPF-12 Facility Name:
V. C.
Summer Inspection Conducted: March 2 C 1993
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Inspector:
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'I Nick Eco'nomos P
,Date Signecf f
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Approved by:
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J/ J eBlake, Chief, Date Signed j
Nate, rials and Process Section
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/ ngineering Branch E
Division of Reactor Safety SUMMARY
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Scope:
This routine, announced inspection was conducted in the areas of inservice inspection (ISI) activities including: eddy current (E/C) examination of steam generator (S/G), tubes, volumetric and surface examination of safety j
related piping welds, reactor vessel closure studs; stuck closure stud
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removal; steam generator replacement project.
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Results:
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At the close of this inspection, preliminary E/C examinations analysis results indicated that approximately 414 tubes in the three S/Gs were candidates for sleeving or plugging. Procedures and examination techniques used by NDE personnel on safety-related welds were satisfactory.
NDE personnel performing surface and volumetric examinations were adequately trained and qualified to perform their assigned tasks; quality records reviewed were satisfactory.
Within the areas inspected violation or deviations were not identified.
9304210019 930416 PDR ADOCK 05000395 O
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l REPORT DETAILS l
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Persons Contacted j
Licensee Employees f
J. Archie, Steam Generator Coordinator f
R. Caban, ISI Coordinator
- R. Clary, Manager Steam Generators Project (SGP)
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- R. Fowlkes, Manager Nuclear Licensing j
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- R. Kelso, NDE Level III Examiner SGP l
- A. Koon, Project Coordinator, Nuclear Operations l
T. Ogburn, Systems Engineer
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T. Prosser, Health Physics Supervisor
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- J. Skolds, Vice President, Nuclear Operation i
- A. Torres, Associate Manager QC i
J. Waters, Erosion / Corrosion Specialist I
Other licensee employees contacted during this inspection included
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technical support, QA and administrative personnel.
Other Organizations
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i Westinghouse Nuclear Services Division (WNSD)
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j R. Keck, Field Service Supervisor l'
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Babcock & Wilcox Nuclear Services Company (BWNSC)
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i R. Pruitt, Site Manager
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T. Richards, NDE Technology Manager i
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1 NRC Resident Inspectors
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- R. Haag, Senior Resident Inspector j
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- Attended exit interview e
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j 2.
Inservice Inspection (ISI)
This is the seventh (7th) refueling outage, and it occurs in the third (3rd,) 40-month examination period of the first (1st) 10-year interval
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of this Unit. The inspector observed in-progress examinations, and i
reviewed procedures and records indicated below, to determine whether ISI examinations were being conducted in accordance with' applicable procedures, regulatory requirements and licensee commitments.
The
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applicable code for examination activities was the American Society of-
Mechanical Engineers Boiler and Pressure Vessel (ASME B&PV) Code, s
Section XI 1977 edition with addenda through Summer 1978.
Ebasco-l Services Inc., under contract with South Carolina Electric and Gas
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Company (SCE&G), was conducting the manual Ultrasonic (UT), Liquid Penetrant (PT), Magnetic Particle (MT), and some Visual (VT)
examinations.
Eddy Current (EC) examinations of the steam generator (
tubing was conducted by Babcock and Wilcox Nuclear Services Co. (BWNS),
l Special Product Inspection Services (SPIs), organization. WNSD, had the l
task of the ten year reactor vessel examination.
a.
Review of NDE Procedures (73052)
The inspector reviewed the procedures listed below to determine whether they were consistent with code requirements and regulatory l
commitments. The procedures were also reviewed in_ the areas of l
procedure approval, requirements for qualification of NDE l
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personnel, visual acuity requirements and compilation of required j
records.
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SCEG/Ebasco ISI Procedures
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SCEG-UT-S78-1, Rev. I Magnetic Particle Examination of Welds and Bolting SCEG-UT-S78-1, Rev. O Ultrasonic Examination of Class 1 and 2 Piping Welds Joining Similar and Dissimilar Materials SCEG-UT-S78-2, Rev. O Straight Bean Ultrasonic Examination
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of piping Welds
SCEG-UT-S78-5, Rev. 1 Ultrasonic Straight Beam Examination l
SCEG-UT-S78-9, Rev. I Ultrasonic Examination of Austenitic Stainless Steel Piping SCEG-VT-S78-1, Rev. I Visual Examination, VT-1 l
SCEG-VT-S78-3, Rev. O Visual Examination for Mechanical l
and Structural Condition of Components and their Supports VT-3&4 SCEG-PT-578-1, Rev.1 Liquid Penetrant Examination b.
Observation of Work Activities (IP73753)
The inspector observed work activities, reviewed certification records of NDE equipment and materials, and reviewed NDE personnel l
qualifications for personnel, utilized for ISI examinations observed. The observations and reviews conducted by the inspector are documented below.
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(1)
Ultrasonic Examination Weld Isometric /
System Comments Drawing No.
W-11 CGE-2-2201A Main Steam 32"x1.150, One spot geometric indication attributed to OD crown configura-tion W-9 CGE-1-4205A CVC 3"sch 80, four spot indications attributed to ID geometry (2)
Magnetic Particle Examinations RPV, Closure Studs 29, 33 and 37 No indications IMS-07-133 RPV identified i
These examinations were performed by personnel who were adequately trained to perform their assigned tasks.
Indications were thoroughly investigated and adequately documented in accordance with procedural requirements.
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Weld number (W-9) above was examined in response to IEB 88-08 Supplement 2, Thermal Stress in Piping Connected to Reactor Coolant Systems, and no evidence of cracking was noted.
The inspector noted that working conditions in the immediate' areas where these examinations took place were less than satisfactory in that:
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A temporary tent had been erected with black plastic material to allow for wet fluorescent magnetic i
particle testing of the reactor vessel closure studs.
Inside the tent, the inspector noted that the closure studs were resting flat just off the floor on wooden.
beams.
In order to perform the required test, each stud, weighing in excess of 600 lbs., had to be rolled manually several times by the technicians, who had to kneel on the floor to accomplish the task.
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addition to performing the task from the position
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described, the inspector noted that the tent had no
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provisions for ventilation to exhaust the kerosene fumes from the aerosol cans used to spray the wet
particles on the surface being examined. This made
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l breathing conditions inside the tent most difficult, requiring the' technicians to stop frequently and exit the tent for fresh air.
Finally, the inspector noted that because of the temporary nature of the tent, stray light entered the enclosure from the floor and the sides thus making testing conditions difficult.
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These conditions were addressed as a concern, during the exit interview and the licensee agreed to look further into this situation to see what improvements
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could be made in this area. During earlier
discussions with cognizant personnel the inspectors
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ascertained that a machine to perform these type tests was on site but had not been installed in the hot l
machine shop for lack of space.
(b)
Weld W-9 above was on the 3-inch charging line attached to reactor coolant loop (RC) piping which had
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a radiation level of 450 MR on contact. The inspector noted that the licensee had made no effort to place
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-temporary protective shielding on the RC piping to i
minimize personnel exposure. As indicated by the
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health physics (HP) survey, the pipe weld being i
examined and the attached RC pipe was sufficiently i
radioactive to set off the warning light on the
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detectors worn by the technicians (a husband and wife
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team) and to a lesser extent, that worn by the inspector. Early on in the inspection, the male j
technician's detector audio alarm activated; he left
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the area and reported to HP.. The female technician
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continued and completed the examination.
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Following the inspection, the inspector questioned the lack of protective lead shielding on the RC piping because it unnecessarily exposed personnel to.a significant dosage (male 73MR, female 67MR), and because this was not a practice commonly observed elsewhere by the inspector.
Through discussions with cognizant licensee personnel, the inspector ascertained that based on HP procedure, HPP-819 Rev. 7, Temporary Shielding Evaluation Installation and Removal, temporary shielding is normally not used unless an exposure savings of at
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least 200 MREM can be achieved. Also, in response to IE Information notice No. 83-64 Lead Shielding Attached to Safety Related Systems Without 10 i
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CFR50.59, the licensee does not permit temporary lead shielding to be placed directly over safety related seismically qualified piping (RC loop), because a 10 CFR 50.59 evaluation to address these conditions has not been done.
i Under the licensee's current program, the only means of providing temporary shielding in these situations is to construct a scaffolding support, upon which l
shielding blankets may be wrapped around. Therefore, the licensee reasons that more worker exposure would i
be realized by building these scaffolds than the amount received by NDE technicians during the examination.
For a more in-depth evaluation of the licensee's program, the inspector requested assistance from the Region II Facilities Radiation Protection Section who is now handling this matter.
Within the areas inspected violations or deviations were not identified.
3.
Eddy Current Examination of Steam Generator Tubes (IP73753)
a.
Background The three S/Gs at V.C. Summer are Westinghouse D-3 models with 4,674 tubes per S/G. The tubing material is mill annealed inconel 600, with 0.750" OD and 0.43" nominal wall thickness. The tubes l
are hard rolled for the full length of the tube sheet on both ends i.e. hot and cold legs. Primary Water Stress Corrosion Cracking (PWSCC) - ID originating cracking in the rolled tube sheet area has been responsible for the majority of plugged tubes at VC Summer.
In general the area of concern is 1-1.5" from the tube
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end to the last main roll expansion at or near the top of the tube l
sheet.
i Peripheral tubes in the pre-heater region of the generator are susceptible to support plate wear. AVB wear in the U-bend area throughout the generators is also present.
Some indications of ODSCC and IFA/ SCC have been reported at tube support plate locations throughout the tube bundle. Most indications have been located at the 02H elevation; however, limited reporting of indications at the 05H elevation have also occurred.
V.C. Summer has operated on All Volatile Treatment, (AVT),
chemistry since start-up. To date sludge accumulation has been l
minimal, <10 pounds of sludge was removed in 1991. The chemical
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analysis of sludge indicated an Fe, (Iron), content between 25% to
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60% and Cu (Copper),
content between 2.5% to 6.0%. The only j
components containing copper in the secondary system are the
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moisture separator reheaters which are being replaced during this outage. The replacements are outfitted with inconel tubing.
b.
Examination Program and Procedures ISI activities during this autage included eddy current examination of steam generator tubing. Data acquisition was performed by B&W Nuclear Services Company under contract with the licensee. This activity was performed in accordance with I
requirements and procedures set forth in the Eddy Current Examination Manual (Manual) for SCE&G, V. C. Summer, Unit 1.
As stated earlier in this report, the applicable code was ASME Section XI (77S78). Code Cases invoked for this examination were N-401 and N-401-1.
Requirements of Regulatory Guide 1.83, Rev.1, July 1975, were also applicable by reference. The scope of this EC examination included the following examinations:
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Full-length bobbin coil examinations of all tubes in Steam Generators A, B, and C.
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Rotating pancake coil probe examinations from 2 inches above the secondary face to 4 inches below the secondary face of the inlet tubesheets for all tubes in Steam Generators A. B.
and C.
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Rotating pancake coil probe examinations of all Inconel 600 plugs in the inlet of Steam Generators A, B, and C.
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Rotating pancake coil probe examination of 19 plugs in the outlet of Steam Generator A. 41 plugs in the outlet of Steam Generator B, and 200 plugs in the outlet of Steam Generator C.
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Rotating cross-wound probe examination of 30 sleeved tubes
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in steam Generator A.
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Nominal size bobbin coil probe gauging for denting of 46 sleeved tubes in Steam Generator A, and 49 sleeved tubes in Steam Generator B.
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Approved E/C procedures reviewed for technical content were:
Number Title Revision I
ISI-424 Multifrequency Eddy Current Examination
of.750" OD x.043" Wall RSG Tubing for ASME Exam, and Wear at Tube Support Plates Isl-460 Technical Procedure for the Evaluation of
Eddy Current Data of Nuclear Grade Steam Generator Tubing 151-510 MRPC/ EDDY - 360 System Operating Procedure
151-511 RSG Sleeve Post Installation Eddy Current
l Examination SGP-005 Stearn Generator Tube Eddy Current Data Analyst Guidelines By review of these documents and through discussions with cognizant licensee personnel the inspector ascertained that Eddy Current data was being evaluated in accordance with the requirements of procedure SGP-005 Rev. O which governed the work l
and superseded possible conflicting requirements in BWNS l
procedures.
Equipment used in data acquisition and analysis included Hewlett-Packard (HP) 700 Series computers, MIZ-18A remote data acquisition units, Zetec Eddynet acquisition modules and software modified by B&W to include manipulators tooling control programs, HP fiber-optic hub with HP optic transreceiver and the HP 650/A optic disks
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drive.
Probes used for the examination included differential
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l bobbin probes 0.610" outside diameter and rotating pancake multi-coil probes.
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l Steam generator tubes scheduled for examination during this outage were as follows:
S/G "A" S/G "B" S/G "C" Standard Bobbin 4293 4036 4120 l
Rotating Pancake 4008 3678 4028 Coil (RPC)
Sleeve Crosswound 147 Special Interest - RPC HL-91 HL-190 Plugs-RPC IR-253 HL-305 H1-257 l
l CL-105 CL-40 CL-20 l
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i As indicated above, the entire length of all S/G tubes in service were examined with multifrequency bobbin coil.- In addition,
specific tube locations, TSP instructions' and the area of interest at the top of the tubesheet (+) 2" to (-) 4" on the hot and cold
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leg sides, as required.
l At the time of this inspection,' standard bobbin examinations were approaching comaletion in all tnree S/Gs and by the end of the j
inspection almost all results had been analyzed including those i
acquired with RPC. Preliminary results provided by the licensee and verified by record review were as follows.
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S/G Tube Status Prior to Present (7th) Refueling Outage:
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"L" Total Plugged Tubes 381 636 554 1573 j
Sleeved Tubes 285 358
735 Before this outage, the total number of plugged tubes (1573),
f expressed in terms of percent was about 11.5%. By analysis the licensee determined that a maximum of 18% of the total number of
tubes (20% in one S/G), could be plugged without exceeding the
licensed plugging limit.. In addition, because the licensee could
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not oredict, prior to the outage, the outcome of the EC -
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l exhMtion or the number of tubes requiring plugging; the
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licen ee requested NRR to authorize the use of the Interim l
Plugging Criteria (IPC), for a single cycle. No decision had been
reached on this submittal at the close of this inspection.
However, preliminary results of the EC examinations, reported at the end of this inspections suggested that the requested IPC i
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authorization would probably not be required. This was based on
results which showed that number of tubes requiring repair, added
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to the total number of tubes already plugged, equaled about 16%,
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which was below the 18% plugging limit. Following the closing of i
this inspection, the inspector ascertained that contrary to
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original plans, the licensee decided against sleeving any tubes i
and had determined the number of tubes to be plugged would be as
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follows:
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Tubes To Be Plugged
"A*
"B"
"C" TOTAL 181 255 *
212 648
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- This figure includes three tubes which will be pulled for
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examination. All three tubes were located near the flow distribution baffle plate on the hot leg side of this S/G. The subject tubes were identified as follows:
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Tube Row Column Voltaae A
20 9.84 B
41 11.59 C
43 22.3 These tubes had been previously examined under Westinghouse guidelines and had been found acceptable.
The inspector reviewed certification records of fifteen EC examiners, seven calibration standards, five MlZ-18A RDAU units, i
quality records of explosive plugs and sleeves on site and,
records of surveillances performed on this activity by the licensee.
Within the areas inspected violations or deviations were not identified.
4.
Steam Generator Replacement Project (SGRP) (IP 37700).
The licensee is making preparations for steam generator replacement
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(SGRP) during the scheduled outage in the fall of 1994. During that
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outage, the present Model D3 steam generators will be replaced with
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Westinghouse model Delta-75 with feed rings. By memorandum to the
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commission dated March 12, 1993, entitled Piping Analyses for the Delta-75 Steam Generator, the licensee announced that a re-analysis is being
performed for major piping connected to the steam generators. The re-
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analysis for the RCS loop and component requalification was being
performed by the licensee in conjunction with Westinghouse.
Re-analysis for the mainsteam, feedwater, the emergency feedwater and blow-down systems is being done by the licensee in conjunction with Bechtel and ABB/Impell. By attachment to the subject letter, the licensee provided a description of the methdologies being used for these analyses and indicated that snubber and pipe whip restraint reduction was also being pursued. These changes were being planned under provisions in 10 CFR50.59. Also, according to the subject document the l
analyses will be performed in accordance with the existing plant design j
basis criteria, outlined in the FSAR.
The design basis piping code is ASME Section III, 1971 Edition through Summer 1973 Addenda. An exception will be taken for the fatigue requalification of Class 1 piping components where the 1977 Edition and Addenda through Summer 1979 Addenda will be used. Also by review of the subject document, the inspector ascertained that the Delta-75 replacement S/G piping nozzle locations have no effect on existing pipe routes of the mainsteam and blowdown piping. However, the new feedwater and emergency nozzles will be approximately thirty-three (33), feet above the existing nozzles. The new emergency feedwater nozzles will be on the same elevation, but rotated counterclockwise from the existing nozzles. Also, through discussions with management, the inspector ascertained that contractors were presently performing feedwater l
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t support-plate inspections, laser templating and engineering walkdowns
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inside containment. The licensee is currently developing certain I
documents, identified as Modification Request Forms (MRF's), which will l
scope out specific tasks and provide guidelines / instructions to control
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replacement activities. Station Administrative Procedure SAP-1001
" Conduct of Steam Generator Project Activities" was reviewed for content. This procedure establishes the program for conducting
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activities relative to the SGRP which include design, inspection, l
procurement, planning, scheduling etc.,
under the licensee's QA i
I program. An engineering report on this project is scheduled for early next year.
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l Within the areas inspected violations or deviations were not identified.
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5.
Removal of Reactor Vessel Closure Stud Stuck in Reactor Vessel Flange, l
(IP62700).
Through discussions and records review the inspector ascertained that
reactor vessel head stud number 43 had seized, in place, after approximately one inch (l") of rotation, during a previous reactor vessel head disassembly. (This occurred on or about October-19,1985.)
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Nonconformance Notice 2070 dated October 19, 1985, was issued with the i
recommended disposition that the stud be left in place. An engineering
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evaluation was to be performed prior to head removal. Westinghouse Safety Evaluation SECL-85-473 Rev. I dated November 8, 1985, stated the following: thread engagement was marginally acceptable with the stud in the "as-is" position (turned out 2.05 turns); thread protection provided
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no corrosion concerns; inspection of the bolt and flange would not be violated since they were to be examined once every ten years, according (
to Section XI requirements. Finally, the safety evaluation concluded
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that leaving the stud in the reactor vessel flange, did not constitute a safety hazard or violate code and regulatory requirements.
During the current outage, the licensee contracted ABB Combustion Engineering Nuclear Services, to remove the subject stud. This activity was to be performed under procedure HMSI-93M3126 Rev. O, Reactor Vessel Stud Removal, Inspection and Repair, dated March 21, 1993. Under this procedure the stud was'to be cut flush with the flange surface, drilled /
machined out to a diameter not to exceed 5.870".
The remaining threads were to be removed using hand tools to collapse the threads and pull them out of the stud hole. At the close of this inspection the inspector ascertained that the stud was removed successfully as planned.
Within the areas inspected violations or deviations were not identified.
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Exit Interview The inspection scope and results were summarized on March 26, 1993, with those persons indicated in paragraph 1.
The inspector described the areas inspected and discussed in detail the inspection results. The licensee did not identify as proprietary any of the material provided to or reviewed by the inspectors during this inspection. Dissenting
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comments were not received from the licensee.
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