IR 05000456/2008008
| ML082880242 | |
| Person / Time | |
|---|---|
| Site: | Braidwood  |
| Issue date: | 10/13/2008 |
| From: | Dave Hills NRC/RGN-III/DRS/EB1 |
| To: | Pardee C AmerGen Energy Co |
| References | |
| IR-08-008 | |
| Download: ML082880242 (20) | |
Text
ber 13, 2008
SUBJECT:
BRAIDWOOD STATION, UNITS 1 AND 2 EVALUATIONS OF CHANGES, TESTS, OR EXPERIMENTS AND PERMANENT PLANT MODIFICATIONS BASELINE INSPECTION REPORT 05000456/2008008(DRS); 05000457/2008008(DRS)
Dear Mr. Pardee:
On September 12, 2008, the U.S. Nuclear Regulatory Commission (NRC) completed the evaluations of changes, tests, or experiments and permanent plant modifications inspection at your Braidwood Station, Units 1 and 2. The enclosed report documents the inspection results, which were discussed on September 12, 2008, with Mr. Larry Coyle 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.
The inspectors reviewed selected procedures and records, observed activities, and interviewed personnel.
Based on the results of this inspection, no findings of significance were identified.
In accordance with 10 CFR 2.390 of the NRCs 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 NRCs document system (ADAMS). ADAMS is accessible from the NRC Website 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. 50-456; 50-457 License Nos. NPF-72; NPF-77 Enclosure: Inspection Report 05000456/2008008(DRS); 5000457/2008008(DRS)
w/Attachment: Supplemental Information cc w/encl: Site Vice President - Braidwood Station Plant Manager - Braidwood Station Regulatory Assurance Manager - Braidwood Station Chief Operating Officer and Senior Vice President Senior Vice President - Midwest Operations Senior Vice President - Operations Support Vice President - Licensing and Regulatory Affairs Director - Licensing and Regulatory Affairs Manager Licensing - Braidwood, Byron and LaSalle Associate General Counsel Document Control Desk - Licensing Assistant Attorney General J. Klinger, State Liaison Officer, Illinois Emergency Management Agency Chairman, Illinois Commerce Commission
SUMMARY OF FINDINGS
IR 05000456/2008008(DRS); 05000457/2008008(DRS); 08/25/2008 - 09/12/2008; Braidwood
Station, Units 1 & 2; Evaluations of Changes, Tests, or Experiments and Permanent Plant Modifications.
The inspection covered a two-week announced baseline inspection on evaluations of changes, tests, or experiments and permanent plant modifications. The inspection was conducted by three regional based engineering inspectors. Based on the results of this inspection, no findings of significance were identified. The significance of most findings is indicated by their color (Green, White, Yellow, Red) using Inspection Manual Chapter (IMC) 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 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-Revealed Findings
Cornerstone: Initiating Events
No findings of significance were identified.
Cornerstone: Mitigating Systems
No findings of significance were identified.
Cornerstone: Barrier Integrity
No findings of significance were identified.
Licensee-Identified Violations
No findings of significance were identified.
REPORT DETAILS
REACTOR SAFETY
Cornerstones: Initiating Events, Mitigating Systems, and Barrier Integrity
1R17 Evaluations of Changes, Tests, or Experiments and Permanent Plant Modifications
.1 Evaluations of Changes, Tests, or Experiments
a. Inspection Scope
From August 25, 2008, through September 12, 2008, the inspectors reviewed 10 evaluations performed pursuant to 10 CFR 50.59 to determine if the evaluations were adequate and that prior NRC approval was obtained as appropriate. The inspectors also reviewed 22 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.
Documents reviewed are listed in the attachment to this report.
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.
This inspection constitutes 10 samples of evaluations and 22 samples of changes as defined in Inspection Procedure 71111.17-05.
b. Findings
During this inspection, the NRC senior resident inspector requested the teams assistance with their review of the 10 CFR 50.59 evaluation EC361637 (FDRP 23-003)
Abandon the Upper Cable Spreading Room Carbon Dioxide (CO2) System, Revision 0.
The results of that review will be documented in the Braidwood Integrated Inspection Report 2008004.
.2 Permanent Plant Modifications
a. Inspection Scope
From August 25, 2008, through September 12, 2008, the inspectors reviewed 13 permanent plant modifications that had been installed in the plant during the last three years. The modifications were chosen based upon risk significance, safety significance, and complexity. As per Inspection Procedure 71111.17, one modification was chosen that affected the design bases and functioning of interfacing systems as well as introducing the potential for common cause failures. The inspectors reviewed the modifications to verify that the completed design changes were in accordance with the specified design requirements, and the licensing bases, and to confirm that the changes did not adversely affect any systems' safety function. Design and post-modification testing aspects were verified to ensure the functionality of the modification, its 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 also used applicable industry standards to evaluate acceptability of the modifications. The list of modifications and other documents reviewed by the inspectors is included as an attachment to this report.
This inspection constitutes 13 samples as defined in Inspection Procedure 71111.17-05.
b. Findings
- (1) Temporary/Permanent Conversion of Lead Shielding on Piping Systems
Introduction:
The inspectors identified an unresolved item (URI) concerning seismic Category I pipe supports associated with the safety injection (SI) and residual heat removal (RH) piping systems. Design documents for the 2SI06 and 2RH01 piping subsystems pipe supports were not sufficiently detailed to demonstrate compliance with the American Institute of Steel Construction (AISC) Manual of Steel Construction Code and the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code.
Description:
The SI and RH piping systems are part of the emergency core cooling system (ECCS). The Braidwood Updated Final Safety Analysis Report (UFSAR),
Section 6.3.1, stated the primary function of the ECCS is to remove the stored and fission product decay heat from the reactor during accident conditions. The ECCS also provides shutdown capability for design basis accidents by means of boron injection.
The ECCS was classified as a safety class II system designed to meet seismic category I requirements.
The inspectors reviewed Calculation BRW-97-0827-M, Piping Evaluation for Lead Shielding Installation on Subsystem 2SI06 Piping per Temporary Lead Shielding Request (TSR) No.95-153, 96-018,96-045, 96-053, and 97-120, Revision 0 and Minor Revision 0A. The purpose of Revision 0 was to evaluate the affect of the temporary lead shielding installed on the 2SI06 piping subsystem (i.e., lead shielding installed on sections of pipelines 2RH01BA-12, 2RH01BB-12, 2RH01CA-16, 2RH01CB-16, 2SI06BA-24, and 2SI06BB-24) by the TSRs. The purpose of Minor Revision 0A was to evaluate the affect of converting the temporary lead shielding to permanent lead shielding. In addition, Minor Revision 0A identified that based on recent industry concerns the lead shielding weighed up to 10 percent more than was previously analyzed.
Pipe stresses were determined from loads and load combinations due to internal pressure, pipe system dead weight, pipe thermal expansion and seismic excitation.
The 2SI06 and 2RH01 piping subsystems were designed to the ASME Boiler and Pressure Vessel Code,Section III, Subsection NC and ND, 1977 Edition up to and including the 1979 Addenda. The associated pipe supports were designed to the AISC Manual of Steel Construction Code and the ASME Boiler and Pressure Vessel Code,Section III, Subsection NF, 1977 Edition through Summer 1979 Addenda. The seismic response spectra analysis of the piping subsystem was analyzed using the ASME Code Case N-411, Alternative Damping Values for Seismic Analysis of Classes 1, 2, and 3 Piping Sections,Section III, Division 1. As specified in the licensees UFSAR, Section 3.7, Table 3.7-1, Damping Values, the ASME Code Case N-411 may be used for alternative damping values when NRC conditions as defined in Regulatory Guide 1.84, Design and Fabrication Code Case Acceptability ASME Section III Division I, Revision 24 are met.
In Calculation BRW-97-0827-M, Revision 0, the licensees qualification of the pipe supports were based on their review of Calculation 13.2.29, Structural Calculation for Mechanical Component Support [Pipe Support Number], Revision 2 and the use of engineering judgment. The licensee concluded that sufficient margin existed in the pipe support design such that the supports would be able to withstand the increased loads from the installed lead shielding.
The inspectors reviewed the Structural Calculation for Mechanical Component Support
[Pipe Support Number] contained in Calculation 13.2.29, Revision 2, for the following:
Pipe Support Number Pipe Support Number Pipe Support Number 2SI06309X 2SI06328X 2SI06342X 2SI06310X 2SI06335X 2SI06345X 2SI06311G 2SI06336X 2SI06351X 2SI06316X 2SI06337X 2SI06358X 2SI06318X 2SI06340S 2SI06360X The inspectors determined that the engineering judgment used in Calculation BRW-97-0827-M, Revision 0 was not valid and the aforementioned pipe supports could exceed their design basis and operability acceptance limits. Also, a condition specified in ASME Code Case N-411 stated This Code Case is not appropriate for analyzing the dynamic response of piping systems using supports designed to dissipate energy by yielding. The licensee initiated issue report (IR) 00816677, NRC MOD/50.59 Inspection - 2SI06 Piping Subsystem Support, dated September 11, 2008, to address this issue.
In response to IR 00816677, the licensees prompt operability determination concluded through analysis that the fillet weld connection between the attachment plate and the embedment plate for pipe support 2SI06316X exceeded design basis and operability limits. Subsequently, the licensee performed a walkdown to field verify the actual fillet weld size of this connection. The actual fillet weld size was determined to be 7/16 thick, which was greater than the 1/4 thick fillet weld size used in the analysis. The licensee determined that the pipe support 2SI06316X fillet weld connection exceeded the design basis limits but was within operability acceptance limits.
Further analysis by the licensee showed pipe supports 2SI06328X, 2SI06340S, 2SI06345X and 2SI06358X exceeded their design basis limits. The concrete expansion anchor bolt evaluation for each pipe support resulted in a factor of safety of less than four but greater than two. A factor of safety of greater than two satisfies the operability requirements specified in procedure OP-AA-108-115, Operability Determinations (CM-1), Revision 6.
Pipe support 2SI06318X was determined to have no margin for design basis acceptance limits and pipe support 2SI06337X had minimal design basis margin remaining. The licensee determined that the pipe supports 2SI066309X, 2SI06310X, 2SI06311G, 2SI06335X, 2SI06336X, 2SI06342X, 2SI06351X and 2SI06360X met design basis and operability acceptance limits.
This issue is considered unresolved pending the licensees response to address the inspectors request for additional information (RAI) regarding the following inspectors concerns:
- The licensees prompt operability determination was made without incorporating the additional 10 percent weight and installed locations of the temporary lead shielding identified by Calculation BRW-97-0827-M, Minor Revision 0A into the licensees pipe stress computer analysis program. These changes have the potential to affect all pipe supports and anchors on the 2SI06 and 2RH01 piping subsystems.
- There are approximately 58 pipe supports and five pipe anchors associated with the 2SI06 piping subsystem, which were not verified to determine their design basis acceptability.
- Pipe supports and anchors for the 2RH01 piping subsystem were never verified to determine their design basis acceptability.
- A condition specified in the ASME Code Case N-411 states When used for reconciliation work or for support optimization of existing designs, the effects of increased motion on existing clearances and on line mounted equipment should be checked. This condition along with the other four ASME Code N-411 specified conditions were not addressed in the pipe stress analysis for either the 2SI06 or 2RH01 piping subsystems.
At the end of this inspection, the licensee stated that the SI and RH piping systems pipe stress computer analysis program would be revised to reflect the RAI concerns and the results provided to the NRC for review within three months of September 12, 2008.
(URI 05000457/2008008-01(DRS))
OTHER ACTIVITIES (OA)
4OA2 Identification and Resolution of Problems
.1 Routine Review of Condition Reports
a. Inspection Scope
From August 25, 2008, through September 12, 2008, the inspectors reviewed corrective action process documents that identified or were related to 10 CFR 50.59 evaluations and 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 team are listed in the attachment to this report.
b. Findings
No findings of significance were identified.
4OA6 Management Meetings
.1 Exit Meeting Summary
On September 12, 2008, the inspectors presented the inspection results to Mr. Larry Coyle and other members of the licensee staff. The licensee acknowledged the issues presented. The inspectors confirmed that none of the potential report input discussed was considered proprietary.
.2 Interim Exit Meetings
No interim exits meetings were conducted.
ATTACHMENT:
SUPPLEMENTAL INFORMATION
KEY POINTS OF CONTACT
Licensee
- L. Coyle, Plant Manager
- C. Furlow, Design Engineering
- G. Golwitzer, Regulatory Assurance
- J. Gosnell, Design Engineering
- D. Gustofson, Design Engineering
- D. Ibrahim, Design Engineering
- J. Knight, Nuclear Oversight
- T. McCool, Operations
- J. Morales, Design Engineering
- R. Gadbois, Maintenance Director
- J. Odeen, Project Management Director
- J. Petty, Regulatory Assurance
- D. Riedinger, Design Engineering Manager
- B. Schipiour, Work Management Director
- M. Smith, Engineering Director
Nuclear Regulatory Commission
- B. Dickson, Senior Resident Inspector
- A. Garmoe, Resident Inspector
- J. Heath, Resident Inspector
Attachment
LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED
Opened
- 05000457/2008008-01(DRS) URI RAI To Determine Adequacy of Pipe Supports Designed for Design Basis Loading Conditions with Lead Shielding Installed on SI and RH Subsystems (Section 1R17.2b.(1))
Closed and
Discussed
None Attachment