IR 05000388/1992027
| ML17157C069 | |
| Person / Time | |
|---|---|
| Site: | Susquehanna  |
| Issue date: | 11/08/1992 |
| From: | Gray E, Patnaik P NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML17157C068 | List: |
| References | |
| 50-388-92-27, NUDOCS 9211180381 | |
| Download: ML17157C069 (8) | |
Text
U. S. NUCLEAR REGULATORY COMMISSION
REGION I
REPORT NO.
DOCKET NO.
LICENSE NO.
LICENSEE:
50-388/92-27 50-388 NPF-22 Pennsylvania Power and Light 2 North Ninth Street Allentown, PA 18101 FACILITYNAME:
INSPECTION AT:
Susquehanna Steam Electric Station, Unit 2 Berwick, Pennsylvania INSPECTION DATES:
October 6-16, 1992 INSPECTOR:
P. Patnaik, Reactor Engineer, Materials Section, Engineering Branch, DRS Date APPROVED BY:
E. Harold Gray, Chief, Materials Section, Engineering Branch, DRS Date SHLLIIN: h i
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fifth refueling outage.
Review of data taken during the outage including disposition of findings.
Also, the erosion/corrosion thickness measurement, data evaluation and repair/replacement of degraded components were witnessed.
~Re ult: The licensee's ISI program meets the requirements of the applicable ASME Section XI Code and the NRC regulations.
The thickness measurement for erosion/corrosion, data evaluation and repair/replacement efforts were in accordance with approved procedures and the specihcation.
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1.0 INSERVICE INSPECTION (ISI) PROGRAM {Inspection Procedure IP 73051)
An inservice inspection program is mandated by the ASME Boiler and Pressure Vessel Code,Section XI and is essential to protect public health and safety, in that, it assures structural integrity and leak tightness of the reactor coolant system pressure boundary.
The current outage (fifth refueling, fall '92) of Susquehanna Unit 2 is the first scheduled refueling outage of the third period of the first inspection interval. Unit 2 willhave another refueling outage before the expiration of the inspection interval.
The applicable code for the inspection interval is the 1980 Edition of the ASME Section XI Code, including all addenda up to winter 1981.
The inspection plan for the third period was reviewed against the ten-year plan to verify that the selected components and their examinations were in compliance with the requirements of the applicable code.
The licensee's inspection plan also covered augmented inspections of systems and components from the ten-year plan.
The review of the inspection plan for the fifth refueling outage did not identify any discrepancy.
The licensee had adequate staff to maintain the ISI program, which encompassed coordination of contractors'ctivities, witnessing inspections, data review, coordinating resolution of findings and record keeping.
The ISI related activities were noted to be well organized and have progressed on schedule.
Allnon-conforming conditions such as flaws and deficiencies discovered during inservice inspection were documented in nonconformance reports (NCRs) and were routed to nuclear plant engineering by quality control for disposition.
The NCRs are tracked by the compliance department for their status and target dates for completion.
Upon completion of recommended rework or repair, quality control verifies the corrective action, accepts the component and closes out the NCR.
For modification within the ISI boundary, the licensee generates a code repair form, specifying nondestructive examination (NDE) required per the applicable ASME code.
The NDE reports are reviewed by the ISI group and the modification gets incorporated into the ISI plan.
The nuclear quality assurance department performed an audit of the ISI program from June 1 to July 13, 1992 (NQA Audit No.92-050).
The audit results revealed the existence of programs and implementing procedures that adequately addressed ASME Code and other licensing commitments.
The quality control and the ISI group routinely performed surveillance of contractors'ctivities during the outag ~
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2.0 INSERVICE INSPECTION - OBSERVATION OF WORK AND WORK ACTIVITIES(IP 73753)
The inspector witnessed ultrasonic examinations being conducted on core spray nozzle to safe-end and safe-end to pipe welds using an automated, computer based, ultrasonic system (GE SMART 2000).
The manual ultrasonic examinations were witnessed for the following welds and components.
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DCA-207-2-FW-10 (Core Spray System)
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DBA-205-1-4-B (RCIC System)
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Reactor vessel bolts examination The personnel performing the examinations were knowledgeable on the examination procedure.
The inspector verified that certifications of all examination personnel were available in the file.
The documentation on calibration blocks for examination of the above welds was reviewed.
The blocks were in compliance with the ASME Code Section V, 1980 Edition. The licensee maintained and controlled the blocks properly as found during the inspection.
3.0 INSERVICE INSPECTION DATAREVIEW AND EVALUATION(IP 73755)
The ultrasonic data on the following welds were reviewed to ascertain whether the licensee's recording, evaluating and disposition of findings are in compliance with the applicable code and the NDE procedures.
Weld Identifi tion
$gstem SPD BA 2171-FW-17 DBB 2212-FW-5 DBB 2212-2-C DBA 2012-FW-10 N9 Nozzle Cap DCA 207-2-FW-10 DBA 205-1-4-B Main Steam RCIC RCIC RWCU Reactor Vessel Core Spray RCIC The review of data was satisfactory.
The disposition of ultrasonic indications in the following components of the reactor internals were reviewed and corrected or found to be acceptable for continued operation for the current fuel cycl hr ud H d Hold D wn Boltin Examination Forty-eight shroud head hold down bolts attached to the steam separator were ultrasonically examined.
The bolt ¹10 showed evidence of cracking in a creviced region created by a 304 stainless steel ring welded to an alloy 600 series inconel shaft.
The licensee replaced the cracked bolt with a new bolt.
et P m Hold Down Beam Ex mination Twenty jet pump hold down beams located on top of the jet pumps between the shroud and the reactor vessel wall were ultrasonically examined.
Only one beam, No. 17, had an indication in the ligament area near the top thread of the beam top surface.
The indication was estimated to be approximately 0.020" in depth.
The licensee plans to monitor the indication for growth during the next refueling outage of the unit.
t m D er Su ort Rin Examinati n
Eight specific areas on the steam dryer support ring which had previous cracks were ultrasonically examined for measurement of crack depth.
The average crack growth in these areas between the last examination and the current examination was 0.074".
The maximum depth of crack at a given location is 0.66".
The licensee plans to monitor the crack growth during the next refueling outage of the unit.
4.0 CONCLUSION Based on review of records and interviews with personnel, the licensee's ISI program meets the requirements of the applicable ASME Code Section XI and the NRC regulations.
From observation of work, the licensee's nondestructive examination activities were found to comply with the requirements of the applicable procedures and the ASME Code Section V.
Licensee's evaluation and disposition of indications are in compliance with the applicable code and procedures.
5.0 EROSION/CORROSION (E/C) MONITORING (49001)
The monitoring of erosion/corrosion in high energy piping is important to maintain structural integrity of piping and components.
During this inspection, thickness measurements in the field were witnessed and licensee's evaluation of thickness data and disposition of degraded components were also reviewe.1 Observation of Work Activities The NRC inspector witnessed ultrasonic thickness measurement on the following components in the high energy systems.
m nen ID
$g~tem SPH BD 2095 E2 SPH BD 2096 E3 SPH BD 2711 E1 SPH BD 2711 E3 GBD 2042 E2 Extraction Steam Extraction Steam Turbine Sealing Steam Steam Seal Evaporator Feedwater Heater Vent The components were gridded for measurement in accordance with the procedure.
The thickness measurements were taken in accordance with procedure GE-IS1-417, Rev. 1.
The following components in the third extraction piping from the low pressure turbine were repaired or replaced due to excessive wall thinning as a result of erosion.
The inspector performed a check on the following repair/replacement activities.
om nen ID HBD 2092-E5 HBD 2092-E6 HBD 2093-ES HBD 2095-E9 HBD 2092-E1 26" diameter 45'lbow 26" diameter 45 elbow (including 14 ft. of straight pipe between the elbows)
Approx. 8'f 18 inch diameter pipe in the IP condenser Approx. 8'f 18 inch diameter pipe in the IP condenser 26 inch diameter, 90'lbow IJIKIO'~eair HBD 2091-E8 HBD 2091-E5 HBD 2093-E8 HBD 2095-E6 Two weld overlays One weld overlay One weld overlay One weld overlay The work in progress was in accordance with the welding procedure.
5.2 Erosion/Corrosion Data Review The data on thickness measurements were reviewed to ascertain whether the licensee's disposition of findings are in compliance with the "General Specification for Inspection of ASME Class 1, 2, 3 and B 31.1 Piping for Moisture Erosion/Corrosion" (Specification M-1414).
The inspector verified the licensee's calculated erosion percentage and evaluated the disposition of certain components.
The inspector accepted the licensee's resolution of findings. It was noted that the third extraction piping had high erosion and ten components in the system were repaired or replaced during the outage.
The components in this system willbe monitored for wall thinning during the next refueling outage.
5.3 Conclusion The licensee's thickness measurement for erosion/corrosion, data evaluation and repair/replacement efforts of degraded components were in accordance with approved procedures and specifications.
The engineering evaluation of components for repair/replacement were thorough and accurate, 6.0 ACTIONON PREVIOUSLY IDENTIFIEDITEMS - (CLOSED) UNRESOLVED ITEMS - DOCUMENTATIONOF ULTRASONIC TESTING DEMONSTRATION (50-387/91-05 AND 50-388/91-05)
An unresolved item was identified in NRC inspection 91-05, due to lack of documentation of demonstration of the Procedure GE-1S1-435, Rev. 0, "Automated Ultrasonic Testing of Reactor Pressure Vessel Assembly Welds," to the authorized inspector as required under ASME Code Section V, Article I, Section T-150.
Based on licensee's review and verification that the procedure had been qualified and demonstrated to the authorized inspector, as stated in the disposition of the NCR, the above unresolved items are closed out.
7.0 ENTRANCE AND EXITMEETINGS Members of the licensee's management, engineering and technical staff were informed of the scope and the purpose of the inspection at the entrance meeting which took place on October 6, 1992.
The findings of the inspection were presented to and discussed with members of the licensee's management at the conclusion of the inspection on October 16, 1992.
A list of attendees of the exit meeting is appended to this report as Attachment A%I'ACHMENTI LIST F ATTENDEES Penn 1v ni P wer Li h R. A. Baker N. T. Fedder E. W. Figard J. T. Lindberg G. Stanley T. K. Steingass D. E. Tillery H. Webb R. Wehry U
Nuclear Re ulat NQA/QC ISI Manager, Nuclear Maint.
ISI Supt. of Plant Maint. Tech.
ANII/FM Maint. Tech.
Compliance mmi ion G. S. Barber P. Patnaik NRC-SRI NRC, Reactor Eng.