IR 05000373/2007007
| ML071550429 | |
| Person / Time | |
|---|---|
| Site: | LaSalle  |
| Issue date: | 06/04/2007 |
| From: | Dave Hills NRC/RGN-III/DRS/EB1 |
| To: | Crane C Exelon Generation Co, Exelon Nuclear |
| References | |
| IR-07-002 | |
| Download: ML071550429 (21) | |
Text
June 4, 2007
SUBJECT:
LASALLE COUNTY STATION, UNITS 1 AND 2 NRC EVALUATION OF CHANGES, TESTS, OR EXPERIMENTS, AND PERMANENT PLANT MODIFICATIONS BASELINE INSPECTION REPORT 05000373/2007007(DRS); 05000374/2007007(DRS)
Dear Mr. Crane:
On May 18, 2007, the U.S. Nuclear Regulatory Commission (NRC) completed baseline inspections of Evaluation of Changes, Tests, or Experiments and Permanent Plant Modifications at the LaSalle County Station. The enclosed report documents the results of the inspection which was discussed with Ms. Susan Landahl and other members of your staff at the completion of the inspection on May 18, 2007.
The inspectors examined activities conducted under your license as they relate to safety and compliance with the Commissions Rules and Regulations, and with the conditions of your license. The inspectors reviewed selected procedures and records, observed activities, and interviewed personnel.
Based on the results of the inspection, two NRC-identified findings of very low safety significance were identified. However, because these violations were of very low safety significance and because they were entered into your corrective action program, the NRC is treating the issues as Non-Cited Violations (NCVs) in accordance with Section VI.A.1 of the NRCs Enforcement Policy.
If you contest the subject or severity of a NCV, you should provide a response within 30 days of the date of this inspection report, with the basis for your denial, to the U. S. Nuclear Regulatory Commission, ATTN: Document Control Desk, Washington, DC 20555-0001, with a copy to the Regional Administrator, U.S. Nuclear Regulatory Commission - Region III, 2443 Warrenville Road, Suite 210, Lisle, IL 60532-4352; the Director, Office of Enforcement, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001; and the Resident Inspector Office at the LaSalle County Station. In accordance with 10 CFR 2.390 of the NRC's "Rules of Practice," a copy of this letter, its enclosure, and your response (if any), will be available electronically for public inspection in the NRC Public Document Room or from the Publicly Available Records (PARS) component of NRC's document system (ADAMS), accessible from the NRC Web site at http://www.nrc.gov/reading-rm/adams.html (the Public Electronic Reading Room).
Sincerely,
/RA/
David E. Hills, Chief Engineering Branch 1 Division of Reactor Safety Docket Nos. 50-373; 50-374 License Nos. NPF-11; NPF-18 Enclosure:
Inspection Report 05000373/2007007(DRS); 05000374/2007007(DRS)
w/Attachment: Supplemental Information cc w/encl:
Site Vice President - LaSalle County Station LaSalle County Station Plant Manager Regulatory Assurance Manager - LaSalle County Station Chief Operating Officer Senior Vice President - Nuclear Services Senior Vice President - Mid-West Regional Operating Group Vice President - Mid-West Operations Support Vice President - Licensing and Regulatory Affairs Director Licensing - Mid-West Regional Operating Group Manager Licensing - Clinton and LaSalle Senior Counsel, Nuclear, Mid-West Regional Operating Group Document Control Desk - Licensing Assistant Attorney General Illinois Emergency Management Agency State Liaison Officer Chairman, Illinois Commerce Commission
SUMMARY OF FINDINGS
IR 05000373/2007002(DRS); 05000374/2007002(DRS); 4/23/2007 - 5/18/2007; LaSalle County
Station, Units 1 and 2; Evaluation of Changes, Tests, or Experiments (10 CFR 50.59), and Permanent Plant Modifications.
The inspection covered a 3-week announced baseline inspection on evaluations of changes, tests, or experiments and permanent plant modifications. The inspection was conducted by three Region based inspectors. Two Green Non-Cited Violations (NCVs) were identified. The significance of most findings is indicated by their color (Green, White, Yellow, Red), using Inspection Manual Chapter 0609, Significance Determination Process (SDP). Findings for which the SDP does not apply, may be Green, or be assigned a severity level after NRC management review. The NRC's program for overseeing the safe operation of commercial nuclear power reactors is described in NUREG-1649, Reactor Oversight Process, Revision 3; dated July 2000.
A.
Inspector-Identified and Self-Revealed Findings
Cornerstone: Barrier Integrity
C
- Green.
The inspectors identified a finding having very low safety significance and an associated NCV of 10 CFR Part 50, Appendix B, Criterion III, Design Control, involving an inadequate maintenance procedure used to remove drywell head bolts. Specifically, in maintenance procedure MA-AB-756-600 Reactor Disassembly, the licensee failed to provide instructions to remove only every other bolt to ensure that the drywell head assembly configuration remained within the analyzed configuration for operating Modes through 3. As a corrective action, the licensee intended to provide additional procedure instructions to restrict bolt removal to every other bolt, or delete the procedure option for early bolt removal with the plant in Modes 1 through 3.
The finding was determined to be greater than minor because absent NRC intervention the inadequate procedure could lead to a more significant problem. Specifically, procedure MA-AB-756-600 would have allowed removal of bolts from adjacent locations on the drywell head assembly which could affect the structural and/or leakage integrity of the containment. The finding was of very low safety significance based on a Phase 1 screening in accordance with IMC 0609, Appendix A, Significance Determination of Reactor Inspection Findings for At-Power Situations, because it did not represent an actual open pathway for containment, and did not involve a reduction in defense in depth for the atmospheric control or hydrogen control function of containment. The primary cause of this finding was related to the cross-cutting area of human performance because the licensee did not provide complete, accurate, and up to date design documentation to plant personnel. (Section 1R17)
C
- Green.
The inspectors identified a finding having very low safety significance and an associated NCV of 10 CFR Part 50, Appendix B, Criterion XII, Control of Measuring and Test Equipment, involving lack of calibrated tools used to establish torque for the drywell head assembly bolts. Specifically, for five air hammer wrenches used to install drywell head assembly bolts on Unit 1 and Unit 2, the licensee failed to ensure these tools were properly calibrated to confirm the accuracy of the torque applied. The licensee entered this issue into the corrective action program, performed an operability evaluation, and concluded that sufficient torque had been applied to the drywell head bolts. The licensee operability conclusion was based upon the vendor advertised torque wrench specifications, torque margins available in the design analysis, and periodic air hammer wrench maintenance.
The finding was determined to be greater than minor because absent NRC intervention the lack of calibration testing for these wrenches could lead to a more significant problem. Specifically, the drywell head assembly bolts may not receive sufficient torque to establish a preload which assures containment leakage and structural integrity. The finding was of very low safety significance based on a Phase 1 screening in accordance with IMC 0609, Appendix A, Significance Determination of Reactor Inspection Findings for At-Power Situations, because it did not represent an actual open pathway for containment, and did not involve a reduction in defense in depth for the atmospheric control or hydrogen control function of containment. The primary cause of this finding was related to the cross-cutting area of human performance because the licensee did not provide adequate and available facilities and equipment (e.g. calibrated equipment)for personnel reassembling the drywell head. (Section 1R17)
Licensee-Identified Violations
None
REPORT DETAILS
REACTOR SAFETY
Cornerstones: Initiating Events, Mitigating Systems, and Barrier Integrity
1R02 Evaluations of Changes, Tests, or Experiments
.1 Review of 10 CFR 50.59 Evaluations and Screenings
a. Inspection Scope
From April 23, 2007 through May 4, 2007, and May 14 through 18, 2007, the inspectors reviewed five safety evaluations performed pursuant to 10 CFR 50.59 to determine if the evaluations were adequate and to determine if prior Nuclear Regulatory Commission (NRC) approval was obtained if applicable. These five safety evaluations were all that the licensee had approved for changes to the plant since completion of the last NRC review of 10 CFR 50.59 safety evaluations. Therefore, the inspectors considered this portion of the procedure completed with less than minimum expected number of safety evaluations (six) as identified in NRC inspection procedure 71111.02. The inspectors also reviewed 18 screenings where licensee personnel had determined that a 10 CFR 50.59 evaluation was not necessary. In regard to the changes reviewed where no 10 CFR 50.59 evaluation was performed, the inspectors verified that the changes did not meet the threshold to require a 10 CFR 50.59 evaluation.
The evaluations and screenings were chosen based on risk significance, safety significance, and complexity. To assess the adequacy of these evaluations and screenings, the inspectors used in part, Nuclear Energy Institute (NEI) 96-07, Guidelines for 10 CFR 50.59 Implementation, Revision 1, to determine acceptability of the completed evaluations and screenings. The NEI document was endorsed by the NRC in Regulatory Guide 1.187, Guidance for Implementation of 10 CFR 50.59, Changes, Tests, and Experiments, dated November 2000. The inspectors also consulted Part 9900 of the NRC Inspection Manual, 10 CFR Guidance for 10 CFR 50.59, Changes, Tests, and Experiments. The list of documents reviewed by the inspectors is included as an attachment to this report.
b. Findings
No findings of significance were identified.
1R17 Permanent Plant Modifications
.1 Review of Permanent Plant Modifications
a. Inspection Scope
From April 23, 2007 through May 4, 2007, and May 14 through 18, 2007, the inspectors reviewed ten permanent plant modifications that had been installed in the plant during the last two years. The modifications were selected based upon risk significance, safety significance, complexity, and several modifications were chosen that affected the Barrier Integrity Cornerstone.
For the modification and design reviews completed, the inspectors performed physical inspections of the following plant components:
C Drywell head assembly (DHA) tools and equipment laydown areas on the Unit 2 Refueling Floor; C
Unit 2 digital electro-hydraulic control equipment installed in the Auxiliary Electric Equipment Room and Control Room; C
Replacement scram discharge volume vent and drain pilot valves in the Unit 1 Reactor Building; C
Fuel priming pumps installed on the 0 and 1A emergency diesel generators; and C
Unit 1 hydraulic control unit rod control management system transponder cards and branch amplifier card configuration as displayed on a licensee mockup.
The inspectors reviewed the modifications to verify that the completed design changes were in accordance with the specified design requirements, the licensing bases, and to confirm that the changes did not adversely affect any systems' safety function. In particular, the inspectors reviewed the post-modification testing to ensure the functionality of associated system, and any support systems. The inspectors also verified that the modifications performed did not place the plant in an increased risk configuration. The inspectors applied NRC regulations and applicable industry standards to evaluate acceptability of the modifications. The list of modifications and other documents reviewed by the inspectors are included as an attachment to this report.
b. Findings
b.1 Inadequate Procedure for Removal of Drywell Head Bolts
Introduction:
The inspectors identified a NCV of 10 CFR Part 50, Appendix B, Criterion III, Design Control, having very low safety significance (Green) involving inadequate maintenance procedure used to remove drywell head bolts.
Description:
On February 26, 2007, the licensee removed 30 of 60 DHA bolts with Unit 2 in Mode 3 in accordance with procedure MA-AB-756-600 Reactor Disassembly.
Specifically, Attachment 2 step 9a of this procedure authorized maintenance staff to remove 50 percent of the drywell head bolts with the Unit in any Mode (e.g., included Modes 1 through 3). However, this procedure step did not provide maintenance staff with instructions to remove only every other bolt to ensure the DHA bolted configuration remained within that analyzed in calculation L-002666 Evaluation of LaSalle Drywell Bolts to Justify Permanent Removal of Bolts. Fortunately, the licensee maintenance supervisor verbally instructed workers to remove every other bolt to ensure the DHA remained within the analyzed configuration. However, the inspectors identified that the procedure was inadequate because it allowed removal of 50 percent of the DHA bolts without regard to bolt locations. Specifically, if adjacent bolts were removed, it would have invalidated the DHA analysis and potentially affected containment leakage integrity under a design basis loss-of-coolant accident condition.
This inadequate procedure guidance appeared to stem from a lack of on-site engineering reviews when the original procedure was approved by the site staff in September of 2000, based upon a corporate generated maintenance procedure. The licensees procedure approval process had not changed since 2000, MA-AB-756-600 had undergone 7 revisions, and this deficient procedure was recently used on the Unit 2 DHA, without licensee staff identifying the deficiency. Therefore, the inspectors concluded that the licensee had opportunities to correct this performance deficiency, and that it represented current plant performance.
Analysis:
The inspectors determined that, failure to establish an adequate procedure to ensure that the DHA would remain within analyzed configurations during reactor maintenance activities, was a performance deficiency warranting a significance evaluation. The finding was determined to be greater than minor, because of the absence of NRC intervention, the inadequate procedure could lead to a more significant problem. Specifically, procedure MA-AB-756-600 would allow removal of bolts from adjacent locations on the DHA which could affect the structural and/or leakage integrity of the Unit 1 and Unit 2 containments.
This finding affected the Barrier Integrity Cornerstone because if left uncorrected, it could affected the integrity of the Unit 1 and 2 containment barrier. The inspectors evaluated the finding using inspection manual chapter (IMC) 0609, Appendix A, Significance Determination of Reactor Inspection Findings for At-Power Situations, Phase 1 screening, and determined that the finding was of Green risk because it did not represent an actual open pathway for containment, and did not involve a reduction in defense in depth for the atmospheric control or hydrogen control function of containment. This finding had a cross-cutting aspect in the area of human performance because the licensee did not provide complete, accurate and up to date design documentation to plant personnel.
Enforcement:
10 CFR Part 50, Appendix B, Criterion III, Design Control, requires, in part, that measures shall be established to assure that applicable regulatory requirements and the design basis, as defined in § 50.2 and as specified in the license application, for those structures, systems, and components to which this appendix applies are correctly translated into specifications, drawings, procedures, and instructions.
Contrary to the above, as of May 16, 2007, the licensee had not adequately translated design basis information for the DHA bolted configuration into maintenance procedure MA-AB-756-600 Reactor Disassembly Revision 7. Specifically, in Step 9a of 2 of procedure MA-AB-756-600, the licensee did not limit removal of DHA bolts in Modes 1 through 3 to every other bolt as analyzed in Calculation L-002666. This procedural deficiency had existed since September 19, 2000, when Revision 0 of corporate procedure MA-AB-RS-6-00600 Reactor Disassembly was approved for site use. The licensee documented this issue in AR 00630337 and intend to provide additional procedure instructions to either restrict bolt removal to every other bolt, or delete the procedure option for early DHA bolt removal. Because this violation was of very low safety significance and was entered into the licensees corrective action program, this violation is being treated as an NCV, consistent with Section VI.A.1 of the NRC Enforcement Policy (NCV 05000373/2007007-01; NCV 05000374/2007007-01).
b.2 Lack of Calibration or Testing for Air Hammer Tools Used for DHA Bolt Installation
Introduction:
The inspectors identified a NCV of 10 CFR Part 50, Appendix B, Criterion XII, Control of Measuring and Test Equipment, having very low safety significance (Green) involving lack of calibrated tools used to establish torque for the DHA bolts.
Description:
For the LaSalle Unit 1 and 2 refueling outages the DHA bolts were originally tensioned with hydraulic bolt tensioners. After September of 2000, the licensee changed this process to use air driven impact wrenches for reassembly of the DHA in accordance with procedure MA-AB-756-601"Reactor Reassembly. These air hammer wrenches were not included in the licensees Maintenance and Test Equipment Control Program and no other calibrated device was used to confirm that adequate torque was applied to the DHA bolts. During reassembly, the 60 DHA bolts are considered tight when the air wrench impacting continues for 10 seconds after rotation of the bolt stops. However, the licensee did not conduct impact wrench torque tests to demonstrate that sufficient torque was produced to establish the bolt preload assumed in the DHA analysis. Specifically, the torque assumed by the licensee in the current DHA analysis (calculation L-002666)was 5000 ft-lbs to ensure 74 kips of bolt preload.
Without calibration tests for these air hammer wrenches, the inspectors were concerned that the DHA bolts may not have been sufficiently torqued to meet the bolt preload assumed in the accident analysis. The inspectors identified the following factors which could result in achieving less than design torque for the DHA bolts: 1) no distinguishing marks existed on air hammer wrench socket, which would challenge maintenance workers ability to determine when the bolt head enclosed by the socket had stopped rotating; 2) after the end of bolt rotation, the 10-second interval to continue tightening was not measured so actual times could be less; 3) the inlet air supply to the hammer could fall below 90 psig required by the wrench vendor to provide for maximum torque output; and 4) the air hammers could have a factory or service induced defect on internal components adversely affecting torque output. To minimize the effect of service induced defects on torque output, the licensee had rebuilt one of the five air hammer wrenches approximately every two years (eight to ten years between rebuilds for a given air wrench). The licensee staff concluded that sufficient torque had been applied to the drywell head bolts to justify containment operability based upon; the vendor advertized specifications that these wrenches would produce 10,000 ft-lbs within 6 seconds after bolt rotation stops, torque margins available in the design analysis, and periodic air hammer maintenance.
The use of uncalibrated torque wrenches on DHA bolts appeared to stem from a lack of on-site engineering reviews when the original procedure was approved by the site staff in September 2000, based upon a corporate generated maintenance procedure. The licensees procedure approval process had not changed since 2000, MA-AB-756-601 had undergone 6 revisions, and the licensee continued to use uncalibrated torque wrenches on the DHA for Unit 1 and 2 outages completed since September 2000. Therefore, the inspectors concluded that the licensee had opportunities to correct this performance deficiency, and that it represented current plant performance.
Analysis:
The inspectors determined that, failure to establish calibration testing for air hammer wrenches used on the DHA bolting to ensure adequate torque was applied, was a performance deficiency warranting a significance evaluation. The finding was determined to be greater than minor, because of the absence of NRC intervention, the lack of calibration testing could lead to a more significant problem. Specifically, the DHA bolts may not receive sufficient torque to establish a preload which assured containment leakage and structural integrity.
This finding affected the Barrier Integrity Cornerstone because if left uncorrected, it could adversely affect the integrity of the Unit 1 and 2 containment barrier. The inspectors evaluated the finding using IMC 0609, Appendix A, Significance Determination of Reactor Inspection Findings for At-Power Situations, Phase 1 screening, and determined that the finding screened as Green because it did not represent an actual open pathway for containment, did not involve a reduction in defense in depth for the atmospheric control or hydrogen control function of containment. This finding had a cross-cutting aspect in the area of human performance because the licensee did not provide adequate and available facilities and equipment (e.g. calibrated equipment) for personnel reassembling the DHA.
Enforcement:
Title 10 CFR Part 50, Appendix B, Criterion XII Control of Measuring and Test Equipment, requires in part that Measures shall be established to assure that tools, gages, instruments, and other measuring and testing devices used in activities affecting quality are properly controlled, calibrated, and adjusted at specified periods to maintain accuracy within necessary limits.
Contrary to the above, as of May 16, 2007, the licensee had not adequately established measures for five air hammer wrenches used to install DHA bolts on Unit 1 and Unit 2, to ensure these tools were properly calibrated and adjusted at specified periods to maintain the accuracy of the torque applied. This deficiency had existed since September 19, 2000, when the licensee authorized use of air hammer wrenches for bolt installation in accordance with Revision 0 of corporate procedure MA-AB-RS-6-601 Reactor Reassembly. The licensee documented this issue in AR 00630431 and was evaluating options for use of calibrated equipment for establishing DHA bolt torque including torque tests for the existing air driven impact wrenches. Because this violation was of very low safety significance and was entered into the corrective action program, this violation is being treated as an NCV, consistent with Section VI.A.1 of the NRC Enforcement Policy (NCV 05000373/2007007-02; NCV 05000374/2007007-02).
OTHER ACTIVITIES (OA)
4OA2 Identification and Resolution of Problems
.1 Routine Review of Condition Reports
a. Inspection Scope
From April 23, 2007 through May 4, 2007, and May 14 through 18, 2007, the inspectors reviewed twenty-six corrective action documents that were related to 10 CFR 50.59 evaluations or permanent plant modifications. The inspectors reviewed these documents to evaluate the effectiveness of corrective actions related to permanent plant modifications and evaluations for changes, tests, or experiments issues. In addition, corrective action documents written on issues identified during the inspection were reviewed to verify adequate problem identification and incorporation of the problems into the corrective action system. The specific corrective action documents that were sampled and reviewed by the inspectors are listed in the attachment to this report.
b. Findings
No findings of significance were identified.
4OA6 Meetings
.1 Exit Meeting
The inspectors presented the inspection results to Ms. Susan Landahl and others of the licensees staff, on May 18, 2007. The inspectors returned proprietary information reviewed during the inspection and the licensee confirmed that none of the potential report input discussed was considered proprietary.
ATTACHMENT:
SUPPLEMENTAL INFORMATION
KEY POINTS OF CONTACT
Licensee
- S. Landahl, Site Vice President
- D. Enright, Plant Manager
- T. Simpkin, Regulatory Affairs Manager
- J. Bashor, Engineering Director
- J. Rommel, Design Engineering Manager
- B. Hilton, Design Engineering
- P. Holland, Regulatory Affairs
Nuclear Regulatory Commission
- D. Kimble, Senior Resident Inspector
- F. Ramirez, Resident Inspector
LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED
Opened and Closed
- 05000374/2007007-01 NCV Inadequate Procedure for Removal of Drywell Head Bolts (Section 1R17.b.1)
- 05000374/2007007-02 NCV Lack of Calibrated Air Wrench for Drywell Head Assembly Bolt Installation (Section 1R17.b.2)
Discussed
None