IR 05000388/1994009
| ML17158A303 | |
| Person / Time | |
|---|---|
| Site: | Susquehanna  |
| Issue date: | 05/18/1994 |
| From: | Carrasco J, Modes M, Patniak P NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML17158A302 | List: |
| References | |
| 50-388-94-09, 50-388-94-9, NUDOCS 9406020147 | |
| Download: ML17158A303 (14) | |
Text
U. S. NUCLEAR REGULATORY COMMISSION
REGION I
DOCKET/REPORT NO:
50-388/94-09 LICENSEE:
FACILITY:
DATES:
Pennsylvania Power and Light Company Susquehanna Steam Electric Station Berwick, Pennsylvania April4-8, 1994, and April 11-15, 1994 INSPECTORS'rrasco, Reactor Engineer Materials Section Division of Reactor Safety Date Prakash Patniak, Reactor Engineer Materials Section Division of Reactor Safety Date APPROVED BY:
Michael C. Modes, Chi f Materials Section Division of Reactor Safety Dat 9406020147 940520 PDR ADOCK 05000388
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"v'nd replacement programs of Class 1, 2, and 3 pressure-retaining components, and the mechanical stress improvement process (MSIP) performed on the reactor coolant system to mitigate intergranular stress corrosion cracking in austenitic stainless steel piping.
julg The in-service inspection gSI), repair, and replacement programs of Class 1, 2, and 3 pressure-retaining components are performed in accordance with Technical Specifications.
The activities related to in-service inspection were coordinated properly, and the licensee exercised good control over contractors'ondestructive examination activities.
During the outage, the MSIP was successfully performed on welds.
The inspector closed Unresolved Item No. 92-05-1 on erosion/corrosion of Unit 1 feedwater tee connection (X-175).
DETAILS 1.0 IN-SERVICE INSPECTION (ISI) (73753-1)
1.1 Purpose and Scope The purpose of this inspection was to verify that the licensee's in-service inspection (ISI) and the repair and replacement programs for Class 1, 2, and 3 pressure-retaining components are in compliance with the Technical Specifications (TS), 10 CFR 50.55(a); the American Society of Mechanical Engineers (ASME) Code,Section XI; and relief requests implemented by the licensee.
The inspectors also followed up on the licensee's request for relief from hydrostatic testing requirements in connection with the replacement of the drain valves on the
"B" loop of the reactor recirculation system.
1.2 Structural Integrity Requirement TS Section 3.4.8 states that the structural integrity of ASME Code Class 1, 2, and 3 components shall be maintained in accordance with Specification 4.4.8., which states that the in-service inspection program for ASME Code Class 1, 2, and 3 components shaH be implemented in accordance with Section XIof the ASME Boiler and Pressure Vessel Code and applicable Unit 2 - 1980 through winter '81 addenda, as required by 10 CFR 50.55(a)(g), except where specific written relief has been granted by the NRC pursuant to 10 CFR 50.55(a)(g)(6)(i).
1.3 In-Service Inspection (ISI) of Class 1, 2, and 3 Preside Retaining Components Compliance with 10 CFR 50.55(a) ASME Code The inspector verified that, by letter, dated February 28, 1985, the licensee submitted the first interval (February 12, 1985, to February 12, 1995) ISI program plan.
The ISI program plan corresponding to Unit 2 is in the In-Service Inspection Technical Document No. IST-206 titled, "In-Service Inspection Program Plan First Ten Year Inspection Program."
The inspectors verified that the NRC has reviewed this program, with relief requests, for compliance to 10 CFR 50.55(a), which embraces the ASME Code Section XI.
The inspector verified the licensee's series of key procedures for implementing the ISI program.
The procedures start with the Nuclear Department Administrative Procedure (NDAP), which details responsibilities; Maintenance Technology (MT) for further breakdown of responsibilities; Nuclear Engineering Instruction Manuals (NIEMs) for scheduling and planning of ISI exams and database control; Maintenance Instruction (Ml) for data control; and Snubber Maintenance (SM) for snubber program as shown belo Pmcehue No.
Pm~re Title NDAP-QA-1608 MT-AD-522 MT-GM-055 MT-AD-595 NEIM-QA-1151 NEIM-00-1152 NEIM-00-1153 MI-II-001 MI-II-002 MI-II-003 MI-II-004 MI-II-005 SM-100-001 SM-100-003 SM-100-004 SM-200-001 SM-200-002 SM-200-003 MT-099-006 ME-ORF-112 In-Service Inspection Procedure Rework, Repair and Replacement of ASME Code Components ASME Section XIVisual Inspection and Disposition Snubber Program (SPS AD-QA-195)
ISI Outage Summary Report Preparation ISI Data Management System (STARS)
ISI Outage Planning and Exam Selection Snubber Functional Testing Data Review Program Surveillance of ISI Contractors ISI Data Review and Approval Storage, Control, and Maintenance of ISI Calibration Standards Snubber Visual Testing Data Review Program Mechanical Snubber Visual Inspection 18 Month Snubber Operating History Review Mechanical Snubber Functional Testing Mechanical Snubber Visual Inspection Mechanical Snubber Functional Testing 18 Month Snubber Operating History Review Snubber Removal and Reinstallation Reactor Pressure Vessel Internals ISI Examination 1.4 Repair and Replacement of Class 1, 2, and 3 Pressure Retaining Components The inspectors reviewed, and discussed with the licensee, the Nuclear Department Procedure MT-AD-522, Revision 2, entitled, "Rework, Repair and Replacement of ASME Code Components."
The licensee stated in the procedure that the ASME Code Program Form will
e
accompany the Work Authorizing Document; and, upon completion, the responsible supervisor or an authorized designee willreview the completed repair form to ensure the code requirements have been satisfied.
The licensee indicated that, when it is determined that a Section XIrepair is necessary, a
Code Repair Program is required by this procedure and an ASME Code Program Form is prepared.
The inspector verified that the procedure outlines key instructions for specific activities necessary to assure the structural integrity of thecomponents.
For example:
the procedure clearly states that repair programs for which Nuclear Engineering concurrence is requested, per Procedure NDAP-QA-1602, willbe forwarded to System Engineering for review and approval. Ifthe repairs involve drawing changes, the changes are made in accordance with Procedure MFP-QA-4002, "Drawings and Drawing Control." Welding is accomplished in accordance with NDAP-QA-1205, "Nuclear Department Welding Program,"
and the Susquehanna Welding and Nondestructive Examination Manual."
For replacements, the inspector verified that the procedure had the necessary instructions to assure the structural integrity of the pressure retaining component.
For Section XI replacements, any necessary drawing changes associated with replacements are made through the use of a proper document control mechanism.
The procedure contains the required ASME Section XI nondestructive post-repair and replacement examinations.
In addition, the post-repair/replacement pressure tests are to be performed in accordance with written Procedure NDAP-QA-0480. This procedure also requires that ASME,Section XI, functional tests be performed following maintenance on pressure retained components.
Based on the review of the procedure that governs the repair and replacement of Class 1, 2, and 3 pressure-retaining components, the inspector found the procedure to be adequate and in compliance with requirements of the ASME Code.
The procedure was also found adequate for control and designates responsibilities for repairs and replacements of ASME Boiler and Pressure Vessel Code components at Susquehanna.
1.5 Review of the ISI Database of Class 1, 2, and 3 Pressure Retaining Components Allthe ISI activity of Class 1, 2, and 3 pressure-retaining components is controHed and monitored by use of the ISI database.
The inspector reviewed and observed a practical demonstration on the versatility of ISI files. The inspector noted that the ISI database's main menu had the ability to add component data, to change component data, to add exam-type data, to change exam-type data, and to update history data.
The inspector verified that the database provided a history for each of the pressure-retaining components in the ISI program.
By looking at the database, the inspector noted that the ISI engineer was able to determine the critical ISI parameters, such as component identification, system, exam-type, the last inspection date, day, interval, and period.
In addition, a cross-reference was given in the database'to the ISI package number, which gave more specific information about a component.
This additional information included calibration, scanning
areas, thickness, and contours.
The licensee indicated that these ISI packages are located in the ISI file room with weld and component histories for all ISI inspection locations.
These packages are also available through the plant's Document Control Center.
The licensee displayed other parts of the database showing specific data for mechanical piping, such as line, ISI P&ID, code class, material upstream and downstream, diameter, nominal wall, design class, access, loop, scaffold.
The inspectors randomly selected a valve bolting, integral attachment, and a mechanical piping component to track down and collect critical data from the database.
The inspector found the key information on these three in the database.
Based on the limited sample, the inspector found that the licensee's ISI database appears to be effective, is consistent with the state-of-the art, and it has the necessary information to ensure an effective and efficient data control system.
The licensee also has adequate maintenance instructions that clearly describe data, the ISI data management system,
"STARS," which is an acronym for Scheduling, Tracking, and Reporting System instruction NEIM-00-1152, Revision 0).
1.6 System Pressure Testing of Class 1, 2 and 3 Pressure Retaining Components To assess the licensee's system pressure testing program, the inspector foHowed up on the licensee's request to use ASME Code Case N-416-1.
During the licensee's initial primary containment entry for Unit 2's sixth refuel outage, two Class 1 drain valves on the reactor recirculation system were found to be leaking.
These two valves were identified by the licensee as needing replacement.
The valves are two-inch nominal pipe-size, with socket-weld connections, and are manufactured by Borg Warner and identified as Nos. 243F051B and 243F052B.
ASME,Section XI, 1980 Edition, including the Winter 1981 Addenda, Subparagraph IWA-4400(a), pressure test, states that, after welding repairs on the pressure retaining boundary, a system pressure test shall be performed in accordance with IWA-5000. Based on this requirement, the licensee stated that a localized hydrostatic pressure test could be performed, but this would include the 2P401B reactor recirculating pump seal.
Based on similar industry events, the licensee believed that the integrity of the mechanical seals willbe compromised, and it willrequire an unscheduled seal replacement.
The licensee also considered a "freeze seal," but concluded that this alternative is impractical due to the existing pipe geometric configuration.
The NRC staff stated the other alternative provided by the code is a system hydrostatic test.
Although this only represents a 10% increase in pressure from system leak test, a significant number of alterations to plant systems must be performed to accomplish the task.
This would place the reactor in an off-normal condition, alter plant systems, and challenge the operator's ability to operate the unit safel On March 24, 1994, the licensee submitted a request to the NRC staff for relief from hydrostatic testing requirements with the following alternative provisions:
AH required pre-and post-weld surface (i.e., liquid dry penetrant) NDE shaH be performed in accordance with the original construction code, ASME,Section III, 1971 Edition, including Addenda through Winter 1972.
2.
Preservice inspection (baseline) of the replacement valve-connecting welds shall be performed in accordance with ASME,Section XI, 1980 Edition through Winter 1981 addenda.
Prior to return to service, a VT-2 visual examination shall be performed in conjunction with a pressure-system leakage test at normal operating pressure, in accordance with the pressure-temperature curves in Technical Specification 3.4.6.1 (IWB-5221).
The NRC staff has reviewed the licensee's submittal and found the relief request and the proposed alternative to code hydrostatic test acceptable.
The inspector had no further questions in this regard.
The inspector noted that system pressure testing engineers have a good understanding of the ASME Code sections related to pressure testing.
Furthermore, it was verified that the licensee complies with 10 CFR 50.55(a), and the ASME Section XI Code requirements for pressure testing through the ISI Technical Document ISI-T-206.0 and Nuclear Department Administrative Procedure NDAP-QA-0480. The ISI Technical Documents provide general Code requirements, including Code Cases and a list of approved Code Relief Requests.
The NDAP-QA-0480 provides specific guidance on Code compliance for pressure testing.
Specific system surveillance procedures and drawings have been developed, based on the in-service documents and the administrative procedures.
These procedures are performed on the periodic or interval basis mandated by the Code via the plant surveillance scheduling program.
During refueling outages, aH periodic pressure tests are scheduled as part of the outage schedule to ensure that any component repair/replacement VT-2 exams are performed with the required system exams.
1.7 Conclusion The licensee's ISI of Class 1, 2, and 3 pressure-retaining components met the criteria established in the technical specifications.
The procedure that governs repair and replacement of Class 1, 2, and 3 components was found adequate and in compliance with requirements established in the ASME Code.
The licensee's submittal for the relief request and proposed alternative to code hydrostatic test was found to be acceptabl.0 IN-SERVICE INSPECTION (ISI) {73753)
2.1 Scope The conduct of in-service inspection using. ultrasonic, magnetic particles, and liquid penetrant examination methods helps to ensure the integrity of the pressure boundary.
During this NRC inspection, the scope of examination for the reactor pressure vessel, as delineated in the 10 CFR 50.55{a){g)(6)(ii)A,"Augmented Examination of Reactor Vessel," was reviewed.
A sampling of in-service inspection data and observation of work activities was performed to ascertain ifthe requirements of the American Society of Mechanical Engineers {ASME)
Code,Section XI, the 10 CFR 50.55(a)(g), and the technical specifications were met.
Also during this inspection, the licensee's procedure for use of the Mechanical Stress Improvement Process (MSIP) to mitigate stress corrosion cracking in austenitic stainless steel piping and the data on MSIPs were reviewed.
An unresolved item of the previous NRC inspection (92-05) on erosion/corrosion was also closed out.
2.2 Findings Effective September 8, )992, new regulations were issued regarding augmented examination of reactor vessels.
As a result of these regulations, all licensees must augment their reactor vessel examinations by implementing once, as part of the in-service inspection interval in effect on September 8, 1992, the examination requirements for reactor vessel shell welds specified in Item B1.10 of examination Category B-A of the ASME Code,Section XI. In addition; all previously-granted relief for Item B1.10, examination Category B-A for the interval in effect on September 8, 1992, was revoked by the new regulation (10 CFR 50.55(a)(g)6a).
For the licensees with fewer than 40 months remaining in the interval on the effective date, such as with Susquehanna Units 1 and 2, deferral of the augmented examination until September 1997 is permissible with the conditions stated in the regulations.
The first ten-year inspection intervals for Susquehanna Units 1 and 2 ends on June 1, 1994.
During the inspection interval, the volumetric examination of the reactor vessel shell welds in examination Category B-A, item No. B 1.10 is as foHows:
CIRCUMFERENTIALWELDS WELD IDENTIFICATION PERCENTAGE COVERAGE
AC 85.6 99.9 LONGITUDINALWELDS WELD IDENTIFICATION PERCENTAGE COVERAGE BA BB BC BD BE BF BG BH BK BM BN BP 97.8 96.7 93.6 94.2 90.4
91.8 81.1 81.1 The examination of the circumferential weld AD and the longitudinal welds BK and BM was restricted due to insulation supports, and a-relief was granted by NRC from the Code requirement.
Although the new regulation revoked these relief requests, the licensee will submit amended relief requests to the NRC in the final revision to the first interval ISI
program plan.
The relief requests willaddress limitations in examination coverage in meeting the augmented reactor vessel examination requirement due to permanent interferences.
The inspector witnessed the automated ultrasonic examination of the inner radius of main steam nozzle N3-D conducted using the General Electric Remote Inspection System (GERIS)
and also the examination of the safe-end-to-nozzle weld on the recirculation discharge nozzle N-2F conducted using the SMART 2000 system.
Both examinations were being conducted in accordance with the licensee's procedures, and the personnel seemed to be well familiar with the techniques used to examine these components.
The following examination data was reviewed, and there were no discrepancies noted:
SYSTEM COMPONENT/WELD IDENTIFICATION EXAMINATION PROCEDURE RESULT Reactor Vessel NIANozzle Safe End Automated Ultrasonic GF IS&40, Rev. 0 Acceptable Residual Heat Removal (RHR)
Mainstcam Mainstcam Mahtstcam Mainstcam Corcsp ray Mainstcam Mainsteam Reactor Coro Isolation Cooling (RCIC)
RCIC High Pressure Coolant Injection (HPCI)
DBB-2071-1-B SPDBA-2161-FW-28 SPDBA-2161-FW-27 SPDBA-2161-FW-24 SPDBA-2161-FW-3 GBB 2092 HWAAthrough 6D GBB-2092 HW-2Athough 2D GBB 2013-3-A VNBB214-17-E-BG2 Gauge bolting VNB214-17-K-B62 valve flango bolting GBB 2151-H 1 hangers DBB 2212-H22 hangers DBB 2212-H15 hangers DBB 2141-821 hangers Automated UT Manual UT Manual UT Manual UT Manual UT Magnetic-Paiticlo (MI)
Visual (VT-1)
Visual (VT-1)
Visual (VT-3)
Visual (VT-3)
Visual (VT-3)
Visual (VT-3)
GE-ISI439, Rev. 0 NUT-2, Rev. 2 NUT-2, Rcv. 2 NUT-2, Rev. 2 NUT-2, Rev. 2 NhGWD-I, Rev. 7 NMTWD-l,Rcv. 7 NMTWD-1,Rcv. 7 NVT-1, Rev. 0 NVT-1, Rcv. 0 NVT-3/4, Rev. 0 NVT-3/4, Rcv. 0 NVT-3, Rcv. 0 NVT-3, Rcv. 0 Acceptable Acccptablc Acceptable Acceptable Acceptable Acceptable Acceptable Acceptable Acceptable Acceptable Acceptable Acceptable Acceptable Acceptable The inspector reviewed the procedure and the technique for the Mechanical Stress Improvement Process (MSIP) used by the licensee to mitigate Intergranular Stress Corrosion Cracking (IGSCC) of stainless steel piping. The MSIP is believed to produce compressive
stresses at the inside diameter of the weld by mechanical means via a hydraulically-actuated mechanical clamp used to contact the pipe on one side of the weld.
The foHowing data, pertaining to MSIP on three welds, was presented for review.
SYSTEM WELD NO.
PIPE SIZE INITIAL (diameter)
PRELOAD PRESSURE HYDRAULIC REQUIRED PRESSURE CHANGE IN PIPE UNIT CIRCUMFERENCE DBB-207-1-24 inches FW-3 (pipe to valve)
DBB-207-1-24 inches 1-B N1A noz to 28 inches safe end 2000 + 500 P>>g 2000 + psig 2000 + psig 21,000 psig Min 0.38, Max 1.00 inch 21,000 psig Min 0.38, Max 1.00 inch 33,900 psig Min 0.31, Max 1.00 inch The inspector found that the change in the circumference of each weld was acceptable, in accordance with engineering specifications.
Each weld was ultrasonically examined following MSIP, and there was no evidence of cracking as a result of the squeeze.
The inspector performed surveillance of the calibration block storage area.
The storage area was adequate, and the general condition of blocks was good, with little evidence of corrosion.
The issuance of calibration blocks was controlled satisfactorily.
The certifications of personnel performing Nondestructive Examinations (NDEs) were readily available.
The qualification of NDE personnel was performed in accordance with NQAP-13.1, "Nuclear Department NDE Training, Qualification, and Certification Program."
During in-service inspection of the steam dryer, a crack of approximately 18 inches long was detected in the horizontal weld connecting the drain channel to the dryer skirt. The licensee determined the crack to be rejectable and generated a nonconformance report for its disposition.
The licensee has planned to repair the crack in the drain channel by welding, prior to returning the unit to service.
2.3 Conclusion The in-service inspection program, during the outage of Susquehanna Unit 2, was found to be in compliance with the applicable ASME Section XI Code and the 10 CFR 50 regulatory requirements.
The activities related to in-service inspection were coordinated properly, and the licensee exercised good control over contractors'DE activities.
The MSIP was successfully performed on welds scheduled during the outag.0 CLOSED UNIOHOLVEDIT%M 50-387/92-05-1 During the sixth refueling of Susquehanna Unit I, the licensee estimated excessive wear in a feedwater tee (X-175) connection and deposited a weld overlay.
While this repair was developed as a part of the engineered resolution to the projected condition in this component by the end of the next operating cycle, this was considered to be a noncode repair by the NRC staff. At the time of the inspection, the feedwater tee (X-175), with the exception of the weld overlay, was in conformance with the ASME code.
The overlay is not expected to have any detrimental effect on functioning of the pipe, and the licensee's erosion/corrosion analysis was noted to be conservative.
Should the actual erosion exceed the projected level, the presence of the weld overlay would be beneficial in maintaining pipe structural integrity.
This item was left as an unresolved item, pending review of licensee's analysis of the examination on erosion/corrosion of Unit I feedwater tee connection.
The inspector reviewed the examination data from the seventh refueling outage and concluded that only nominal erosion/corrosion had occurred at location X-175 and the erosion is relatively uniform and consistent with data obtained at other locations in the same system.
Hence, theapplication of a weld overlay was notrequired.
Nevertheless, itreduced the possibility of a pipe leak during the next operating cycle.
Since the overlay is not expected to have any detrimental effect on the pressure boundary, this unresolved item is closed.
4.0 ENTRANCE AND EXITMEETINGS Members of the licensee's management and engineering staff were informed of the scope and purpose of the inspection at the entrance meeting, which took place on April4, 1994.
The findings of the inspection were presented to, and discussed with, members of the licensee's management at the exit meetings conducted on April 8 and April 15, 1994.
The licensee concurred with the findings of the inspection.
A list of attendees at the exit meetings on April 8, 1994, and April 15, 1994, is attached to this report as Attachment I.
Attachment:
Attendees
ATTACHMENT1 ATH<ADEES Exit Meeting of April8, 1994 nn ni r
Li D. Brophy F. Butler T. Dalpiaz J. Finnegan G. Jones G. Kuczynski J. Lindberg R. Linden C. Myers H. G. Stanley T. Steingass H. Webb R. Wehry Project Engineer, Nuclear Systems Engineer Manager, Nuclear Systems Engineering Manager, Nuclear Maintenance Supervisor, Compliance Vice President, Nuclear Engineering Manager, Nuclear Plant Services Nuclear Maintenance Engineer, ISI Inservice Inspection Specialist, ISI Manager, Nuclear Regulatory Affairs Vice President, Nuclear Operations Supervisor, ISI Supervisor, Maintenance Technology Power Production Engineer, Compliance Exit Meeting of April15, 1994 Tom Clymer T. C. Dalpiaz Norman T. Fedder John Finnegan John Lindberg Randy T. Linden Tim Steingass Herb Webb Rick Wehry NQA Coordinator Manager-Nuclear Maintenance ISI Specialist Supervisor-Compliance ISI-Project Engineer ISI Specialist Supervisor-Testing Maintenance Supervisor-Maintenance Technician PP&L Compliance Engineer Nuclear Re ulat mmissi n David J. Mannai NRC Resident Inspector