IR 05000352/2008010
| ML083650028 | |
| Person / Time | |
|---|---|
| Site: | Limerick  |
| Issue date: | 12/29/2008 |
| From: | Doerflein L Engineering Region 1 Branch 2 |
| To: | Pardee C AmerGen Energy Co, Exelon Nuclear |
| References | |
| IR-08-010 | |
| Download: ML083650028 (20) | |
Text
ber 29, 2008
SUBJECT:
LIMERICK GENERATING STATION - EVALUATIONS OF CHANGES, TESTS OR EXPERIMENTS AND PERMANENT MODIFICATIONS TEAM INSPECTION REPORT 05000352/2008010 AND 05000353/2008010
Dear Mr. Pardee:
On November 20, 2008, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at Limerick Generating Station. The enclosed inspection report documents the inspection results, which were discussed on November 20, 2008 with Mr. C. Mudrick, Site Vice President, and other members of your staff.
The inspection 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.
In conducting the inspection, the team reviewed selected procedures, calculations and records, observed activities, and interviewed station personnel.
Based on the results of this inspection, no findings of significance were identified.
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 the NRC's document system (ADAMS). ADAMS is accessible from the NRC Web site at http://www.nrc.gov/reading-rm/adams.html (the Public Electronic Reading Room).
Sincerely,
/RA/
Lawrence T. Doerflein, Chief Engineering Branch 2 Division of Reactor Safety Docket No: 50-352, 50-353 License No: NPF-39, NPF-85 Enclosure: Inspection Report 05000352/2008010 and 05000353/2008010 w/Attachment: Supplemental Information
Mr. Charles
SUMMARY OF FINDINGS
IR 05000352/2008010, 05000353/2008010; 11/03/2008 - 11/20/2008; Limerick Generating
Station; Engineering Specialist Plant Modifications Report The report covers a two week inspection of the evaluations of changes, tests, or experiments and permanent plant modifications. It was conducted by three region-based engineering inspectors. No findings of significance were identified. The NRCs program for overseeing the safe operation of commercial nuclear power reactors is described in NUREG-1649, Reactor Oversight Process, Revision 4, dated December 2006.
NRC-Identified and Self-Revealing Findings
No findings of significance were identified.
Licensee-Identified Violations
None.
ii
REPORT DETAILS
REACTOR SAFETY
Cornerstones: Initiating Events, Mitigating Systems, Barrier Integrity
1R17 Evaluations of Changes, Tests, or Experiments and Permanent
Plant Modifications (IP 71111.17)
.1 Evaluations of Changes, Tests, or Experiments (24 samples)
a. Inspection Scope
The team reviewed five safety evaluations to determine whether the changes to the facility or procedures, as described in the Updated Final Safety Analysis Report (UFSAR), had been reviewed and documented in accordance with 10 CFR 50.59. In addition, the team determined whether Exelon had been required to obtain NRC approval prior to implementing the changes. The team interviewed plant staff and reviewed supporting information, including calculations, analyses, design change documentation, procedures, the UFSAR, the Technical Specifications, and plant drawings, to assess the adequacy of the safety evaluations. The team compared the safety evaluations and supporting documents to the guidance and methods provided in Nuclear Energy Institute (NEI) 96-07, "Guidelines for 10 CFR 50.59 Evaluations," as endorsed by NRC Regulatory Guide 1.187, "Guidance for Implementation of 10 CFR 50.59, Changes, Tests, and Experiments," to determine the adequacy of the safety evaluations.
The team also reviewed a sample of nineteen 10 CFR 50.59 screenings, applicability determinations and equivalency evaluations for which Exelon had concluded that no safety evaluation was required. These reviews were performed to assess whether Exelon's threshold for performing safety evaluations was consistent with 10 CFR 50.59.
The sample of issues that were screened out included design changes, temporary alterations, procedure changes, and setpoint changes.
The five safety evaluations that were reviewed were the only safety evaluations Exelon had performed during the time period covered by this inspection (i.e., since the last modifications inspection). The screenings and applicability determinations were selected based on the risk significance of the associated structures, systems, and components (SSCs).
In addition, the team compared Exelon's administrative procedures, used to control the screening, preparation, review, and approval of safety evaluations, to the guidance in NEI 96-07 to determine whether those procedures adequately implemented the requirements of 10 CFR 50.59. The reviewed safety evaluations, screenings, and applicability determinations are listed in the attachment.
b. Findings
No findings of significance were identified.
.2 Permanent Plant Modifications (10 samples)
.2.1 Feedwater Flow Conversion Constant Adjustment, Unit 1
a. Inspection Scope
The team reviewed a modification that changed the Unit 1 feedwater flow venturi differential pressure to flow conversion constants. The modification installed a new constant for the three feedwater flow orifices. The constants were determined based on a comparison between multiple secondary plant indications, indicated feedwater flow and reactor power. As a result of the modification, indicated feedwater flow was raised to match secondary plant indications. The review was performed to verify that the design bases, licensing bases and performance capability of the feedwater flow indication had not been degraded by the modification. Additionally, the equivalent 10 CFR 50.59 screen associated with this modification was reviewed as described in section 1R17.1 of this report.
The team assessed whether the modification was consistent with assumptions in the design and licensing bases related to accuracy of the feedflow indication. The team reviewed the methodology used to establish the feedflow indication accuracy, as well as the inputs used in establishing the basis for the accuracy of the methodology. The resulting constant adjustments for the flow venturi were reviewed to evaluate whether they were technically appropriate and consistent with the UFSAR. The team verified whether calculations, analysis, procedures, and the UFSAR were properly updated with revised design information. The team also reviewed the design, testing and calibration programs to determine if the revised calibration techniques had been reviewed by Exelon in accordance with their quality assurance design procedures. Finally, the team conducted interviews with engineering staff to verify whether the indicated feedflow maintained the accuracy assumed in the design and licensing assumptions. The documents reviewed are listed in the attachment.
b. Findings
No findings of significance were identified.
.2.2 Abandoned Loose Parts Monitor System, Unit 2
a. Inspection Scope
The team reviewed a modification that removed the loose parts monitor from service.
The modification removed acoustic monitoring devices, changed alarm indications in the control room and left some equipment, such as wiring, retired in place. The review was performed to verify that the design and licensing bases had not been inadvertently changed by the modification. Additionally, the equivalent 10 CFR 50.59 screen associated with this modification was reviewed as described in section 1R17.1 of this report.
The team assessed whether the removal of the monitor was consistent with the design and licensing bases related to the requirements to ensure foreign material in the reactor vessel would not damage structures, systems or components (SSC) such as fuel rods.
The team reviewed the BWR owners group Topical Report and the associated NRC safety evaluation that discussed removal of the equipment to determine if assumptions in the documents were consistent with Unit 2 design and operation. The team also verified that the equipment retired in-place did not affect the operability of other SSCs required to respond to design basis events. The team verified that drawings, calculations, procedures, and the UFSAR were properly updated with revised design information, and verified that applicable operating guidance had been changed. Finally, the team conducted interviews with engineering staff to verify whether other SSCs impacted by the change would continue to function in accordance with the design assumptions. The documents reviewed are listed in the attachment.
b. Findings
No findings of significance were identified.
.2.3 Installation of HPCI Suppression Pool Level Transmitter Head Chambers, Unit 2
a. Inspection Scope
The team reviewed a modification that installed head chambers on the high pressure coolant injection (HPCI) suppression pool level transmitter indication system. The modification was performed as a corrective action to eliminate the cause of previous calibration problems with the instruments and was designed to prevent water from entering the dry portion of the instrumentation tubing. The chambers are used to aid in the calibration of the suppression pool level indicating equipment. The review was performed to verify that the design bases, licensing bases, and performance capability of the level indication not been degraded by the modification. Additionally, the equivalent 10 CFR 50.59 screen associated with this modification was reviewed as described in section 1R17.1 of this report.
The team assessed if the modification was consistent with requirements in the design and licensing bases. This review included assessing if the component safety classification had been maintained, specifically the structural adequacy of the design.
Design assumptions were reviewed to evaluate whether they were technically appropriate and consistent with the UFSAR. The team also verified whether drawings, calculations, calibration instrumentation sheets, and calibration procedures were properly updated based on the new equipment configuration. The team reviewed the post-modification testing to verify that the suppression pool level transmitter system was working properly. Additionally, the team conducted interviews with engineering staff to verify whether the affected SSCs functioned in accordance with the design assumptions, and to verify if the modification corrected the previously identified calibration problems.
The documents reviewed are listed in the attachment.
b. Findings
No findings of significance were identified.
.2.4 Replacement of Residual Heat Removal Testable Check Valve HV-51-1F050B, Unit 1
a. Inspection Scope
The team reviewed a modification that replaced the residual heat removal (RHR) system testable check valve with a new valve. The modification was performed as preventive maintenance to address concerns with wearing of the seating surfaces on the installed valve. The new valve also included modifications from the original design that had been previously evaluated under the Exelon 50.59 process. The review was performed to verify that the design bases, licensing bases, and performance capability of the RHR system had not been degraded by the modification. Additionally, the equivalent 10 CFR 50.59 screen associated with this modification was reviewed as described in section
1R17 .1 of this report.
The team assessed whether the replacement valve was consistent with the design and licensing bases, including whether the welds required to install the valve in the RHR system were performed in accordance with the American Society of Mechanical Engineers (ASME) piping classification requirements. Design assumptions used for fabrication were reviewed to evaluate whether they were technically appropriate and consistent with the UFSAR. The team verified whether drawings, calculations, analyses, and the UFSAR were properly updated with the revised design information. The team also reviewed the post-modification testing to verify the testing was in accordance with ASME Section XI requirements, and to verify that test results appropriately justified system operability. Finally, the team conducted interviews with engineering staff to determine if the affected SSCs would function in accordance with the design assumptions. The documents reviewed are listed in the attachment.
b. Findings
No findings of significance were identified.
.2.5 Replacement of 5-Micron Elements for Control Rod Drive Water Filters, Units 1 and 2
a. Inspection Scope
The team reviewed a modification regarding the replacement of the control rod drive (CRD) system discharge filters AF204 and BF204. The modification replaced the existing 20-micron filters with 5-micron filters to improve reactor recirculation pump seal performance. The review was performed to verify that the design bases, licensing bases and performance capability of the CRD system had not been degraded by the modification. Additionally, the equivalent 10 CFR 50.59 screen associated with this modification was reviewed as described in section 1R17.1 of this report.
The team assessed whether the modification was consistent with the design and licensing bases related to the CRD system. The review included determining if the component safety classification and flow requirements for CRD system were maintained.
Supporting mechanical calculations and analyses for the instrumentation setpoints and maximum differential pressure were reviewed to ensure design limits were not exceeded. The design assumptions associated with the analyses were reviewed to evaluate whether they were technically appropriate and consistent with the UFSAR. The team verified whether drawings, calculations, analyses, procedures, and the UFSAR were properly updated with revised design information and operating guidance. The team reviewed the post-modification testing and conducted interviews with engineering staff to verify the affected SSCs would function in accordance with the design assumptions. The documents reviewed are listed in the attachment.
b. Findings
No findings of significance were identified.
.2.6 Replacement of RCIC Relief Valve PSV-050-1F018, Unit 1
a. Inspection Scope
The team reviewed a modification regarding the replacement of the barometric condenser/lube oil cooler relief valve PSV-050-1F018, with an alternate relief valve design. The review was performed to verify that the design bases, licensing bases and performance capability of the reactor core isolation cooling (RCIC) system had not been degraded by the modification. Additionally, the equivalent 10 CFR 50.59 screen associated with this modification was reviewed as described in section 1R17.1 of this report.
The team assessed whether the replacement relief valve was consistent with assumptions in the design and licensing bases. The team reviewed supporting mechanical calculations and analyses to assess if the component safety classification, relief capacity, seismic analysis, material equivalency were maintained in accordance with the design. Assumptions used in the analyses were reviewed to evaluate whether they were technically appropriate and consistent with the UFSAR. The team verified whether drawings, calculations, analyses, procedures, and the UFSAR were properly updated with revised design information and operating guidance. The team reviewed the post-modification testing to verify whether the affected SSCs would function in accordance with the design assumptions. In addition, the team interviewed the responsible design engineers and walked down the relief valve to detect possible abnormal installation conditions. The documents reviewed are listed in the attachment.
b. Findings
No findings of significance were identified.
.2.7 Revision of Analysis for Post-Accident Heat Sources and Sinks, Units 1 and 2
a. Inspection Scope
The team reviewed changes to the analysis for the emergency service water (ESW)minimum flow required for the residual heat removal (RHR) room coolers. The analysis revised post-accident heat sources and heat sinks which were used as an input to determine the ESW minimum acceptable flow to the RHR room coolers. The review was performed to verify that the design and licensing bases, and performance capability of the ESW or RHR systems had not been degraded by the revised analysis and subsequent modification. Additionally, the equivalent 10 CFR 50.59 screen associated with this modification was reviewed as described in section 1R17.1 of this report.
The team assessed selected design inputs within the analysis to ensure that the minimum credited ESW flow to the various RHR room coolers would maintain each of the rooms at an acceptable temperature per the RHR room equipment environmental qualification specifications. Design assumptions were reviewed to evaluate whether they were technically appropriate and consistent with the UFSAR such that minimum required flows were achieved. The team verified whether calculations, analyses, procedures, and the UFSAR were properly updated with revised design information and operating guidance. The team reviewed the post-modification testing to verify whether the affected SSCs would function in accordance with the design bases assumptions and plant operating procedures. Finally, the team conducted interviews with engineering staff to verify whether the affected RHR room coolers would function in accordance with the design assumptions. The documents reviewed are listed in the attachment.
b. Findings
No findings of significance were identified.
.2.8 Replacement of Emergency Diesel Generator K1 Relay, Unit 1
a. Inspection Scope
The team reviewed a modification to replace the K1 relay in emergency diesel generator (EDG) D-14 subsequent to a failure of the relay that was experienced during surveillance testing. The new relay design was installed because the original relay was no longer available. The review was performed to determine whether the design bases, licensing bases, and performance capability of the relay had been degraded by the replacement.
Additionally, the equivalent 10 CFR 50.59 screen associated with this modification was reviewed as described in section 1R17.1 of this report.
The team assessed whether the modification was consistent with assumptions in the design and licensing bases related to the EDG voltage regulator circuitry. The team reviewed licensing and design documents to determine the component safety classification, auxiliary contactor functional requirements, and seismic requirements for the K1 relay mounting adapter plate. The team evaluated whether the design requirements were maintained in the supporting evaluations and analyses, and whether they were consistent with the UFSAR. The team reviewed selected evaluations, drawings, analyses, procedures, and the UFSAR to determine whether they had been properly updated with revised design information. The team evaluated the post-modification test to determine whether the K1 relay would function in accordance with design requirements. In addition, the team interviewed the responsible design and system engineers to discuss the K1 relay design requirements, surveillance test failure, extent-of-condition evaluations, and evaluation of the subsequent replacement. The documents reviewed are listed in the attachment.
b. Findings
No findings of significance were identified.
.2.9 Replacement of Emergency Service Water Gate Valve 011-1010, Unit 1
a. Inspection Scope
The team reviewed a modification to replace the ESW gate valve due to excessive seat leakage from the valve that was currently installed. The review was performed to determine whether the design bases, licensing bases, and performance capability of the ESW system had been degraded by the modification. Additionally, the equivalent 10 CFR 50.59 screen associated with this modification was reviewed as described in section 1R17.1 of this report.
The team assessed whether the modification was consistent with assumptions in the design and licensing bases related to the ESW system. This review included assessing if the component safety classification and specific safety function were maintained.
Additionally, the technical basis for the allowable leakage was reviewed to evaluate if operability of the ESW system was maintained. The team reviewed design assumptions in the supporting evaluations and analyses to determine whether they were technically appropriate and consistent with the UFSAR. The team reviewed selected drawings, analyses, procedures, and the UFSAR to determine whether they were properly updated with any revised design information. The team evaluated the post-modification test to verify that the appropriate ASME code acceptance criteria were applied to verify leak-tightness and weld acceptance. In addition, the team interviewed the responsible design and system engineers to discuss the modification and the design requirements. Finally, the team walked down the ESW valve to detect any potentially abnormal installation conditions. The documents reviewed are listed in the attachment.
b. Findings
No findings of significance were identified.
.2.10 Replacement of Motor Operator for A Reactor Feedwater Inboard Valve, Unit 2
a. Inspection Scope
The team reviewed a modification to replace the motor operator on the A reactor feedwater inboard maintenance isolation valve (2F011A). The review was performed to determine whether the design bases, licensing bases, and performance capability of the feedwater system had been degraded by the modification. Additionally, the equivalent 10 CFR 50.59 screen associated with this modification was reviewed as described in section 1R17.1 of this report.
The team assessed whether the modification was consistent with assumptions in the design and licensing bases related to the feedwater system. The team reviewed licensing and design documents to determine the component safety classification, scoping applicability for primary containment isolation function, and adequacy of motor electrical characteristics. The team reviewed design assumptions in the supporting evaluations and analyses to determine whether they were technically appropriate and consistent with the UFSAR. The team reviewed selected drawings, analyses, procedures, and the UFSAR to determine whether they were properly updated with any revised design information. The team evaluated the post-modification test to verify that the motor and valve would successfully perform their functions. In addition, the team interviewed the responsible design and system engineers to discuss the modification and the design requirements. The documents reviewed are listed in the attachment.
b. Findings
No findings of significance were identified.
OTHER ACTIVITIES
4OA2 Identification and Resolution of Problems (IP 71152)
a. Inspection Scope
The team reviewed a sample of condition reports associated with 10 CFR 50.59 and plant modification issues to determine whether Exelon was appropriately identifying, characterizing, and correcting problems associated with these areas. The team also evaluated whether the planned or completed corrective actions were appropriate to address the deficiencies. In addition, condition reports written on issues identified during the inspection were reviewed to verify adequate problem identification and incorporation of the problem into the corrective action system. The specific corrective action documents that were sampled and reviewed by the team are listed in the attachment.
b. Findings
No findings of significance were identified.
4OA6 Meetings, including Exit
The team presented the inspection results to Mr. C. Mudrick, Site Vice President, and other members of Exelon's staff on November 20, 2008. The team verified that this report does not contain proprietary information.
ATTACHMENT
SUPPLEMENTAL INFORMATION
KEY POINTS OF CONTACT
Licensee Personnel
C. Mudrick Site Vice President
E. Callan Plant Manger
- R. Dickinson Director, Engineering
- P. Gardner Director, Operations
- R. Kreider Manager, Regulatory Assurance
- R. Harding Manager, Regulatory Assurance
- S. Bobyock Manager, Plant Engineering
LIST OF ITEMS
OPENED, CLOSED AND DISCUSSED
None