IR 05000457/1999008
| ML20195J383 | |
| Person / Time | |
|---|---|
| Site: | Braidwood  |
| Issue date: | 06/14/1999 |
| From: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | |
| Shared Package | |
| ML20195J380 | List: |
| References | |
| 50-457-99-08, 50-457-99-8, NUDOCS 9906180192 | |
| Download: ML20195J383 (14) | |
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U.S. NUCLEAR REGULATORY COMMISSION REGIONlil Docket No:
50-457 License No:
NPF-77 Report No:
50-457/99008(DRS)
Licensee:
Commonwealth Edison Company Facility:
Braidwood Generating Station, Unit 2 Location:
RR #1, Box 84 Braceville,IL 60407 Dates:
April 15-16, 28-29; May 3, 5-7,13,18,1999 Inspectors:
J. Schapker, Reactor Engineer B. Metrow, Illinois Department of N.,
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- Safety
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Approved by:
J. M. Jacobson, Chief, Mechanical Engineering Branch Division of Reactor Safety
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9906100192 990614 PDR ADOCK 05000457 G
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EXECUTIVE SUMMARY Braidwood Generating Station, Unit 2 NRC Inspection Report 50-457/99008(DRS) -
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This inspection included 'an announced review of the inservice inspection (ISI) and steam generator inspection program.' Specifically, the inspection focused on the ISI nondestructive examinations activities, Quality Control activities to assure the ISI program and steam I
generators meet applicable ASME Code and regulatory requirements.
ISI program requirements were implemented in accordance with Regulatory and
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American Society of Mechanical Engineers (ASME) Code requirements. The corporate SG and self assessment was self critical and thorough. (Section M1.1)
Observation of examinations in progress, review of data packages and personnel
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qualifications indicated that the ISI examinations were conducted in accordance with
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applicable procedures and ASME Code. ISI inspection implementation was found to be well-controlled and monitored by licensee engineering and corporate nondestructive
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examination specialist. (Section M1.2)
The Flow Accelerated Corrosion program was aggressive and utilized the latest industry
guidelines and procedures. (Section M1.3)
The IMensee has developed a comprehensive steam generator inspection program.
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Performance of eddy current examination data acquisition, data analysis, and data resolution of the SG tubing was performed to well formulated procedures and guidelines.
(Section M1.4)
The inservice inspection procedures and documentation compiled with the ASME Code
and Technical Specification requirements.- (Section M3)
The knowledge and performance of the engineering staff and contractors in the area of
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ISI was good. Strong corporate and station engineering experience in the steam generator inspection and repair program resulted in effective inspection and maintenance programs. Training for SG analysists was comprehensive and thorough.
(Section M4)
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Report Details
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i 11. Maintenance M1
. Conduct of Maintenance M1.11 ~ Inservios insoection (ISI) Proaram Review (73051. 50001)
a.
, inspection Scope This was the first outage in the second ISI interval at Braidwood 2, and the~ seventh refueling outage (A2R07). The inspector reviewed the outage plan, the method to select some of the planned examinations, and relief requests and interviewed personnel within :
the engineering group responsible for completion of examinations and tests within the ISI program.
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~ The inspector interviewed personnel within the Nuclear Oversight group responsible for
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implementing the station Master Assessment Plan. The inspector reviewed assessment L
documents including procedures, personnel qualifications, assessment schedules, detailed assessment plans, checklists, and notes taken during ongoing assessments of various activities related to the ISI program.
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Observations and Findinas
The licensee and contractor personnel implemented the ISI program in accordance with -
the 1989 edition (no addenda) of ASME,Section XI. The licensee procedures have been revised to address 1989 Section XI requirements. The inspector reviewed the j
current outage ISI scope for Class 1 and 2 components. Of the 129 individual
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components selected for examination, the licensee selected 19 Class 1 piping welds for examination under their Risk Informed-Inservice Inspection (Rl-ISI) program.
q For Braidwood Unit 2, the licensee implemented a partial scope RI-ISI program, limited to Class 1 pipe welds except socket welds. This represents 723 welds that are part of four systems: Chemical and Volume Control System, Reactor Coolant System,
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Residual Heat Removal, and Safety injection.' The program is being developed as an alternative to the Section XI requirements as' permitted in Code Case N-560, which the NRC has not yet approved. Based on a consequences evaluation by Duke Engineering and a degradation mechanisms evaluation by Structural Integrity Associates (sal), an Electric Power Research Institute (EPRI) Rl-ISI team conducted the inspection element selection process in accordance with N-560. Their report, dated April 2,1999 (SIR-99 044), was reviewed by the inspector. The sal report addressed many of the required attributes of an RI-ISI program that are outlined in Code Case N-560, Regulatory Guide 1.178 (issued for trial use September 1998), and NUREG-0800,
. Standard Review Plan, Chapter 3.9.8 (also issued for trial use September 1998). Of the 723 welds under the old Section XI program,88 fell into high risk regions of the matrix (categories 1,2, and 3),341 fell into medium risk (categories 4 and 5) and 294 welds fe!I into Ic,w risk (categories 6 and 7).
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The licensee plans to submit the RI-ISI plan for NRC approval after the Safety Evaluation Report (SER) is received on the generic EPRI RI-ISI technical report, currently scheduled for September 30,1999. For the current outage,38 welds had been identified under the Section XI program as subject for examination. Of these,19 were selected
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using RI-ISI as high risk and were examined. If the RI-ISIis not approved as submitted, l
the next refuel outage (A2R08) is within the current ISI period for Braidwood 2 ar'd the 19 medium and low risks welds not examined during the current outage could be added
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to the scope of A2R08 and examined at the end of period 1.
j The inspector reviewed submittals to NRC of the ISI Plan and requests for relief including revisions thereto. For this outage the licensee had planned on implementing relief requests numbers R-7, R-15, R-17, and R-20. Relief requests R-15 and R-17 were approved by SER in a letter from NRR dated October 26,1998.
Relief request R-20 was a rewrite of Interval 1 relief NR-09, which was approved via SER in an NRR letter, dated October 4,1991. It addresses the ability to achieve full coverage of the 3xamination zone surrounding the reactor pressure vessel (RPV)
head-to-flange weld. However, NRR approval of relief R-20 had not been received.
During the outage, the licensee ultrasonically tested as much of the weld as possible (using the same methods as were used during the Interval 1) and achieved approximately 80 percent coverage. Examination was from both flange and head sides.
Note that although Section XI does not permit deferral directly for item B1.40, RPV head-to-flange weld, Note 4 does permit flange side only examination during periods 1 and 2 with head side examination during period 3. If the NRC does not approve relief R-20, the licensee will have the balance of time until the end of interval 2 during which to develop and implement a supplemental examination technique in order to achieve more zone coverage.
The final relief, R-7, was a resubmittal of two Interval 1 relief requests: NR-12 and NR-23. However, NRR had not yet approved R-7 so the licensee decided to not
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implement it at this time. It involved ultrasonic testing (UT) of the RHR heat exchange nozzle to chell welds. The licensee requested UT of one inlet and one outlet nozzle to shell weld section instead of all four. Additionally, liquid penetrant testing (LPT) and visual examination (VT-2) is performed on all four welds since any crack initiation would be expected to initiate on the outer surface. The licensee deferred performing the tJT examinations until next outage. The licensee did perform PT and VT-2 leakage examinations as was required to fulfill provisions of Interval 1 relief request NR-23.
The inspector reviewed assessment activities conducted since the beginning of the year when the licensee switched from audit to assessment by their Nuclear Oversight group.
Assessment of ISI and flow accelerated corrosion (FAC) was accomplished under two broad areas: engineering and maintenance. Within the scope of engineering assessment, ISI and FAC were two activities that were to be evaluated. During interviews of the lead assessor, the inspector determined there had been no assessment of ISI, FAC or their NDE contractors conducted under the new assessment program.
The inspector reviewed a procedure for conduct of ISI, NSP-ER-3016 revision 2, issued by corporate Nuclear Generation group on October 27,1998. This procedure provides a curis,;.upproach to the implementation of ASME Code and NRC requirements for lSI of Class 1,2,3, MC and CC components and systems. Corporate Nuclear Oversight is currently performing an implementation assessment of this common procedure across all the licensee's stations. In January 1999 Braidwood station was assessed. The results of this implementation assessment indicated the ISI program was complying with the corporate procedure.
Within the scope of maintenance assessment, station maintenance and contractor activities were evaluated. Special processes were part of the evaluation. Both in-house i
and contracted NDE and welding was evaluated. The inspector reviewed Nuclear Oversight assessment 20-99-026, which was ongoing during the current outage. The purpose of this assessment was to evaluate compliance with and verify effectivenss of the applicable licensee NDE, welding and station work procedures. Activities eduated by the assessors included replacement of the reheater vent lines and failed components i
as part of the FAC program. The inspector reviewed the assessment plan, qualifications of assessors and technical specialists, assessors' notes, and follow-up actions.
The inspector also reviewed the self-assessment for the licensee's steam generator program. The assessment was performed by licensee corporate and staff engineering
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and included a wide variety of program components, such as steam generator inspection, degradation assessment, integrity assessment, repair limits, maintenance and repair, primary-to-secondary leakage monitoring, primary and secondary water chemistry, foreign material exclusion and maintenance of secondary side integrity. The assessment was self critical and identified several program improvement needs; corrective action recommendations were addressed and assigned to the appropriate discipline.
Conel sions c.
y The RI-ISI program for Class 1 piping welds other than socket welds was being developed along currently-available industry, ASME and NRC guidelines. Although some of these guidelines were available for trial use or were not formally approved, the licensee has another outage in the current ISI period into which examinations could be performed if approval is not granted.
I Corporate Nuclear Oversight's maintenance assessment was an ambitious program
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using qualified auditors or technical specialists. The level of detail found in the
assessors' notes, and Problem Identification Forms generated, demonstrated that a
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rigorous assessment was being performed.
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' M1.2. : Inservice inspection Observations -
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Insoection Scooe (73753. 737550
- The inspector observed various nondestructive examinations (NDE) of ISI Program -
components during the outage. ' The inspector reviewed NDE data and indication reports generated during the outage. ~
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Observations and Findinos The inspector observed portions of the following Ultrasonic UT exams:
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Line #
lienLE Description Code item 2CS-01 -
6,12~
Elbow-Pipe'
C5.11 2CS-01 15'
Pipe - Pipe C5.11
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- Pipe - Valve 2CS001B C5.11
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C5.12
= 2PZR-01 9A PRZ Shell Long Seam B2.12 2RC-05 3,4,6,7
- . Pipe - Pipe 89.11 2RC-05
.8 Pipe - 14" Nozzle Branch B9.11 -
Elbow-Valve 2Sl8812A C5.11 The inspector observed portons of the following Liquid Penetrant (LPT) exams:
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Description Code item 2CS-04 1:
Pipe-Valve 2CS001B C5.11 2CS-04
Pipe Long Seam C5.12 2RC-05 3,4,6,7
~ Pipe - Pipe.
B9.11 2RC-05
Pipe - 14" Nozzle Branch B9.11
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The inspector reviewed selected data packages including linearity and calibration data, examination data, percentage coverage calculations, NDE summary repods, and indication evaluation reports.. The inspector determined that the methods, extent and
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techniques of examination complied with the ISI program and applicable NDE procedure.
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The results were within acceptance criteria as outlined in the applicable NDE procedure and ASME Code.' NDE results from prior examinations were used to compare with current results. The recording, evaluating and dispositioning of indications complied with the applicable procedure. Licensee Level lli staff conducted training of contractor examiners on new NDE requirements stemming from RI-ISI, and reviewed all data including indication reports and results from RI-ISI examinations.
The inspector reviewed qualifications of contractor personnel performing NDE under the ISI program. Certification records for consumable materials and serial numbers for equipment used during NDE were also reviewed.. The inspector reviewed drawings and material certifications for calibration blocks used during the outage. All documents reviewed above were acceptable.
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i The inspector reviewed videotape of reactor vessel, in Vessel Visual Inspection, i
recorded during the outage. The video was clear and cc itained before and after exam
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documentation of ability to discem a 1/64 inch black line on an 18 percent neutral gray i
card. Videotape included instrument and control rod dris e nozzles, penetrations and -
funnels, the inner head surface, reactor flange surface, hot leg nozzles, keyway lugs, and lower core plate. No anomalies were noted on the awasponding VT report.
During examination under the. head, a piece of duct tare was found stuck on a control rod drive mechanism (CRDM) funnel. A PlF was writ'.en to document this. It was believed the tape came loose from the reactor head stand during setup of remote video equipment, became entwined with cabling, and eventually attached onto the funnel during equipment maneuvers. The tape was removed and the area cleaned. One of the stitch / locking welds for thermocouple connection TC 75 was found cracked. Photos -
were taken and a PIF was written. Since the other weld on the connection had no cracks and the plugs were in place, licensee engineering dispositioned the crack acceptable.
The weld will be inspected next outage (A2R08) to assure the connection has not degraded further.
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Conclusions Observation of examinations in progress, review of data packages and personnel qualifications indicated that the examinations were conducted in accordance with applicable procedures and ASME Code requirements. ISI activities continue to be well-controlled and monitored by licensee ISI engineering and corporate specialist.
M1.3 Review of Flow Accelerated Corrosion (FAC) Program a.
Insoection Scooe (49001)
The inspector interviewed licensee Engineering Programs staff conceming outage scope, selection process, data acquisition methods, analyses techniques and repairs of
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Observations and Findinas There were 210 components identified initially for examination under the FAC program.
Prior to the shutdown, profile radiography was performed on 54 components in the 3 inch reheater vent system. Thinning beyond minimum thickness was detected in 11 components, so the licensee planned replacement of a major portion of the reheater vent system. All of the 3 inch reheater vent system found above the 451' turbine deck elevation except for valves was replaced. This area was selected for its accessibility and close proximity to personnel walkways along the 451' elevation. For the next outage, the station plans to replace the portion below the turbine deck since more preparation and planning for extsnsive scaffolding is necessary.
Of the original scope of 210, this left 156 components to be examined during the outage.
However, twenty-seven components were cut out due to the reheater vent piping
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replacement and thus were not examined. This left 129 components 3 inch and greater in diameter to be tested by UT during this outage. Of these 129, three eroded areas of
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p, extraction steam piping were detected. These resulted in replacement of two 8 inch -
'45 degree ~ elbows and one 12x8 reducer. No sample expansion due to the three failures occurred because all similar ceiriponents in _ sister trains were already included in the initial scope and were being exar.,ined. The replacements were FAC-resistant Crome-Moly carbon steel material. The licensee replaced all pipe material including
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Prior to the outage the licensee updated the CheckWorks computer program, for
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component selection, to revision F. The licensee continues to use CECalc to analyze UT data.~ The inspector reviewed analyses from the current outage.
The inspector observed UT data being taken on the following components:
Drawing #
Rf Description 2EC-FW-10 2FW-244 & 245
.10" pipe to elbow welds The inspector observed a per*Jon of the replacement of the 3 inch reheater vent system
- piping. Observations included welding of welds 88 and 88A under work package 990008810-01. Review of the work package documents including the weld history-record, the weld procedure specificatkm, procedure qualification record, welder qualifications, and material certifications, c.
Conclusions
' The FAC program implementation indicated a proactive approach. Performing profile RT
. prior to ?lant shutdown reduced the large scope by some 25 percent and confirmed the need for prompt replacement of some components near well-traveled areas of the plant.
M1.3 Steam Generator (SG) inspection-a.
Instw*lan Scooe (73753. 73755. 50001)
The inspector observed portions of the eddy current examination (ET) of the Unit 2 SG's, reviewed acquisition and data analyst qualifications, observed secondary side visual examinations and reviewed ET data, safety analysis report and documentation of ET inspection results.
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Observations and Findings The licensee has incorporated the recommendations of Nuclear Energy institute 97-06 Steam Generator Program and the EPRI guidelines in their SG inspection program.
Braidwood Unit 2 SG's are Westinghouse Model D-5 preheater, vertical U-bend type tubes.. Previous tube degradation has been limited to Anti-Vibration Bar (AVB) wear, with some loose parts and top of the tube sheet (TTS) circumferential indications. The
TTS indications were previously plugged and staked. Subsequent tube pulls from Byron Unit 2 SG's found that the TTS indications were not degradation due to intergranular stress corrosion cracking but were due to manufacturing marks or
scratches.
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ET of the SG tubes this outage included:
100 percent examination of the SG's tuoing using bobbin coil.
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25 percent examination of TTS using the plus point probe.
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25 percent of row 1 & 2 U-bends using the plus point probe.
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20 percent motorized pancake coil of Preheater Expansions.
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25 percent of all hot leg dents and dings exceeding 5 volts.
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In addition visual examination of the secondary side and sludge lancing was performed.
The inspector observed the ET contractor (Westinghouse) perform acquisition and resolution analysis of the ET data and reviewed ET data graphics and resolution analyst reports. In addition to the resolution analyst, the licensee employed the services of an independent Quality Data Analyst as an oversight analyst. Included in his responsibility to review ET data, was to assure the acquisition system essential variables were correct (daily), and review of the analyst performance tracking system to monitor the analyst performance.
No TTS indications were detected this outage. Six tubes were plugged due to AVB wear. The new occurence and depth of AVB wear has decreased in the recent inspections. Due to the AVB defective tubes, SG B,C & D were classified as C-2 and SG-A was C-1 with no defective tubes. No additional inspection sampling was required as the initial sample was 100 percent. No other degradation mechanism was detected during the ET examination.
The inspector observed visual examination of the seventh support plate and top of tube sheet area. One foreign object which was identified the previous outage was observed to be located in the same spot, lodged between two tubes. Attempts to retrieve the foreign object during the previous outage were unsuccessful, as. it was lodged firmly between the two tubes. These tubes were ET examined with no degradation or wall thinning identified. The licensed performed an updated safety evaluation in accordance with 10CFR50.59 to leave the foreign object in place. The inspector reviewed the safety evaluation and concluded the basis for not plugging the affected tubes and leaving the foreign object in place was conservative. The affected tubes are identified as Previous Outage No Change in the ET data records, which ensures these tubes and the foreign material continues to be examined and the condition evaluated.
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Conclusions ET was performed conservatively in accordance with ASME Code and regula'ary requirements. The licensee adopted the latest industry SG inspection and maintenance guidelines. The Westinghouse model D-5 SG's performance and reliability appears to be good.
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't M3 Maintenance Procedures and Documentation '
M3.1 ISI Procedure Review a.
Insoection Scope (73052)
The inspector reviewed the following procedures related to NDE:
Procedure /Rev Dale Descriotion -
NDT-C-2/ 25 April 1999 Rl-ISI changes to UT of Pipe Welds NDT-C-38/ 5 February 1999 UT of Cast Stainless Steel Welds NDT-C-45/ 4 February 1999 UT of Pipe from 1.5" to 4" Diameter NDT-C-55/ 2.
February 1999 UT of Welds Refracted Longitudinal NDT-C-63/ 2 November 1997 Referencing, Stamping, Surface Prep NDT-C-75/ 0 '
February 1999 Rx Head Stud UT at Byron /Braidwood NDT-E-2 January 1999 Multi-frequency Eddy Current Acquisition of SG Tubing at Byron /Braidwood NDT-E-3 January 1999 Evaluation of ET Data for SG Tubing at Braidwood/ Byron
- SG Eddy Current Analysis Guidelines, revision 2 dated January 27,1999.
ET Site Specific Performance Demonstration Program (SSPD) Revision 0, January 27, 1999.
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Observations and Findinas
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The inspectors reviewed the above procedureu to ASME Code 1989 edition. No discrepancies were identified. The station's Authorized Nuclear Inservice inspector (ANil) reviewed and approved the referenced procedures. The SSPD procedure was especially well, formulated and followed the latest industry guidelines, c.
Conclusions The NDE procedures were thorough, descriptive, current with industry guidelines and met applicable NRC regulatory and ASME Code requirements.
M4 Maintenance Staff Knowledge and Performance a.
Jmpection Scooe (50001. 73753)
The inspector interviewed the SG engineers, program manager, several ET analysts and observed their work, b.
Observations and Findinas The inspactor observed the ET examination and informally interviewed several analysts involved in resolution of ET data. The analysts were knowledgeable and experienced in the examination requirements. All of the analysts were required to pass the site specific performance demonstration examination which includes the acquisition / analyst procedures and the SG Eddy Current Data Analyst Guidelines. All analysts were (
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required to be certified as a Quality Data Analyst Level llA per the EPRI PWR Steam Generator Examination Guidelines, Appendix G. The licensee's SG engineering group at the site and corporate office worked together to implement a conservative inspection program to assure reliability of the SG's.
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ConclusiQDE The inspector concluded that the knowledge and performance of the ET analysts were -
good. The corporate and site SG engineers form a knowledgeable and dedicated team assuring quality inspection and maintenance of the SG's.
M5 Maintenance Staff Training and Qualification M5.1 ISI Personnel Qualific uons a.
Insoection Scooe (73753)
The inspector reviewed ISI personnel qualifications of licensee and contract personnel performing the ISI activities observed in Section M1.
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Observations and Findinas Qualification of personnel performing NDE work was verified. NDE personnel were knowledgeable of procedural requirements and proficient in the performance of NDE.
Personnel performing NDE were found to have proper qualifications which had been reviewed and accepted by the licensee staff and the ANil.
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NDE was performed by qualified personnel.
V. Management Meetings X1 Exit Summary Meeting
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The inspector presented the inspection results to the licensee management at the conclusion of the inspection on May 18,1999. The licensee acknowledged the findings presented and did not identify any of the report input discussed as proprietary.
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> PARTIAL LIST OF PERSONS CONTACTED
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Commonwealth Edison ',
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- M.' Cassidy, NRC Coordinato;'
D. Chrzanowski, ISI Coordinator -
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T. Luke, Engineering Manager
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' W. McDonough,' SG Program Manager G. Schwartz, Station Manager -
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h H. Smith,' NDE Program Manager l T Sumpkin, Regulatory Assurance Manager -
T. Tulon, Senior W:e President -
R. Wegner, Operatons Manager Westinghouse j
S.' Armbrister, Site Nuclear Manager.
Illinois Department of Nuclear Safety J. Roman, Resident inspector Hartford Insurance and insoection Aaency q
L. Malabanan, ANil A S. Nuclear Reaulatory Commission
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J. Adams, Resident Inspector D. Pelton, Resident inspector C. Phillips, Senior Resident inspector INSPECTION PROCEDURES (IP) USED IP49001:
Erosion / Corrosion Program Inspection -
IP50001:
Steam Generator Program inspection IP73753:
Inservice Inspecticn Observation.
IP73755-Inservice Inspection Review of Data IP73051:
Inservice Inspection Program Review IP73052:
Inservice inspection Procedure Review
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ITEMS OPEN, CLOSED, AND DISCUSSED Ooened None
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Closed None Discussed
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LIST OF ACRONYMS USED ANil'
Authorized Nuclear Inservice inspector -
AVE Anti Vibration Bar ASME'
American Society of Mechanical Engineers CRDM Control Rod Drive Mechanism EPRI Election Power Research Institute ET Eddy Current Examination FAC Flow Accelerated Corrosion
.IP inspection Procedure ISI inservice inspect:on LPT Liquid Penetrant Examination NCR Nonconformance Report NDE Nondestructive Examination NRR, Nuclear Reactor Regulation PDR Public Document Room RI-ISI Risk Informed-Inservice inspection RPV
' Reactor Pressure Vessel sal Structural Integrity Associates SER Safety Evaluation Report SG-Steam Generator
'SSPD Site Specific Performance Demonstration TTS Top of Tube Sheet -
UT Ultrasonic Examination
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