IR 05000456/2011008

IR 05000456-11-008, 05000457-11-008; 09/12/2011 - 09/30/2011, on Braidwood Station Units 1 and 2; Evaluation of Changes, Tests, or Experiments and Permanent Plant Modifications

ML11301A260
Person / Time
Site:	Braidwood
Issue date:	10/28/2011
From:	Robert Daley Engineering Branch 3
To:	Pacilio M Exelon Generation Co, Exelon Nuclear
References
IR-11-008
Download: ML11301A260 (21)

Text

ctober 28, 2011

SUBJECT:

BRAIDWOOD STATION, UNITS 1 AND 2 EVALUATION OF CHANGES, TESTS, OR EXPERIMENTS AND PERMANENT PLANT MODIFICATIONS BASELINE INSPECTION REPORT 05000456/2011008; 05000457/2011008 (DRS)

Dear Mr. Pacilio:

On September 30, 2011, the U.S. Nuclear Regulatory Commission (NRC) completed an Evaluation of Changes, Tests, or Experiments and Permanent Plant Modifications inspection at your Braidwood Station, Units 1 and 2. The enclosed inspection report documents the inspection results which were discussed on September 30, 2011, with Ms. A. Ferko 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, two NRC-identified findings of very low safety significance were identified. The findings involved a violation of NRC requirements. However, because of their very low safety significance, and because the issues were entered into your corrective action program, the NRC is treating these issues as Non-Cited Violations (NCVs) in accordance with Section 2.3.2 of the NRC Enforcement Policy.

If you contest the subject or severity of any 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 Braidwood Station. In addition, if you disagree with the cross-cutting aspect assigned to any finding in this report, you should provide a response within 30 days of the date of this inspection report, with the basis for your disagreement, to the Regional Administrator, Region III, and the NRC Resident Inspector at the Braidwood 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 System (PARS)

component of NRC's Agencywide Documents Access and Management System (ADAMS),

accessible from the NRC Web site at http://www.nrc.gov/reading-rm/adams.html (the Public Electronic Reading Room).

Sincerely,

/RA/

Robert C. Daley, Chief Engineering Branch 3 Division of Reactor Safety Docket Nos. 50-456; 50-457 License Nos. NPF-72; NPF-77

Enclosure:

Inspection Report 05000456/2011008; 05000457/2011008; w/Attachment: Supplemental Information

REGION III==

Docket No: 50-456; 50-457 License No: NPF-72; NPF-77 Report No: 05000456/2011008; 05000457/2011008 Licensee: Exelon Generation Company, LLC Facility: Braidwood Station, Units 1 and 2 Location: Braceville, IL Dates: September 12 - 30, 2011 Inspectors: J. Bozga, Reactor Inspector (Lead)

J. Gilliam, Reactor Inspector M. Jones, Reactor Inspector Approved by: R. Daley, Chief Engineering Branch 3 Division of Reactor Safety Enclosure

SUMMARY OF FINDINGS

IR 05000456/2011008, 05000457/2011008; 09/12/2011 - 09/30/2011; Braidwood Station

Units 1 and 2; Evaluation of Changes, Tests, or Experiments and Permanent Plant Modifications.

This report covers a two-week announced baseline inspection on evaluation of changes, tests, or experiments and permanent plant modifications. The inspection was conducted by Region III based engineering inspectors. Two NRC-identified Green findings were identified by the inspectors. Both findings were considered as Non-Cited Violation (NCV) of NRC regulations.

The significance of most findings is indicated by their color (Green, White, Yellow, Red) using Inspection Manual Chapter (IMC) 0609, Significance Determination Process (SDP). The cross-cutting aspects were determined using IMC 0310, Components Within the Cross-Cutting Areas. 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: Mitigating Systems

• Green: The inspectors identified a finding of very low safety significance (Green) and associated Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for the failure to properly evaluate the structural steel embedment plate which supports Safety Injection (SI) pipe supports 1SI06025V and 1SI06030S. Specifically, the licensee failed to demonstrate compliance with the American Institute of Steel Construction (AISC) and Seismic Category I linear elastic requirements. The licensee entered this issue into their corrective action program and planned calculation revisions and modifications as needed to restore design margins.

The finding was determined to be more than minor because the finding was associated with the Mitigating Systems cornerstone attribute of design control and affected the cornerstone objective of the availability, reliability, and capability of the SI piping and pipe supports. Specifically, the licensee used the actual material yield stress to ensure the structural steel embedment plate would maintain structural integrity when subjected to design loads. This is contrary to the AISC and Seismic Category I linear elastic requirements to use the specified minimum yield stress of the material. The inspectors determined that the finding was of very low safety significance because the finding did not result in loss of operability or functionality. The inspectors did not identify a cross-cutting aspect associated with this finding because it was associated with a calculation from the 1980s and was not reflective of current performance. (Section 1R17.2.b.(1))

Green.

The inspectors identified a finding of very low safety significance (Green) and associated Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for the failure to properly evaluate the Unit 1 SI subsystem 1SI06 and the Unit 1 Chemical Volume and Control System (CVCS) subsystem 1CV18 piping and pipe supports. Specifically, the licensee failed to demonstrate compliance with the AISC and the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code for the 1SI06 and 1CV18 piping and pipe supports. The licensee entered this issue into their corrective action program and planned calculation revisions and modifications as needed to restore design margins.

The finding was determined to be more than minor because the finding was associated with the Mitigating Systems cornerstone attribute of design control and affected the cornerstone objective of the availability, reliability, and capability of SI piping and pipe supports and CVCS piping and pipe supports. Specifically, the licensee did not perform an analysis to ensure compliance with AISC and ASME Section III requirements with the addition of permanent lead shielding to ensure the 1SI06 and 1CV18 piping and pipe supports would maintain structural integrity when subjected to design basis loads. The inspectors determined that the underlying finding was of very low safety significance because the finding did not result in loss of operability or functionality. The inspectors did not identify a cross-cutting aspect associated with this finding because this was a calculational deficiency that did not occur within the past three years and was not reflective of current performance. (Section 1R17.2.b.(2))

Licensee-Identified Violations

No violations of significance were identified.

REPORT DETAILS

1R17 Evaluation of Changes, Tests, or Experiments and Permanent Plant Modifications

.1 Evaluation of Changes, Tests, or Experiments

a. Inspection Scope

From September 12, 2011 through September 30, 2011, the inspectors reviewed six safety 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 15 screenings where licensee personnel had determined that a 10 CFR 50.59 evaluation was not necessary. The inspectors reviewed these documents to determine if:

• the changes, tests, or experiments performed were evaluated in accordance with 10 CFR 50.59 and that sufficient documentation existed to confirm that a license amendment was not required;

• the safety issue requiring the change, tests or experiment was resolved;

• the licensee conclusions for evaluations of changes, tests, or experiments were correct and consistent with 10 CFR 50.59; and

• the design and licensing basis documentation was updated to reflect the change.

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 constituted six samples of evaluations and 15 samples of changes as defined in IP 71111.17-04.

b. Findings

No findings of significance were identified

.2 Permanent Plant Modifications

a. Inspection Scope

From September 12, 2011 through September 30, 2011, the inspectors reviewed 11 permanent plant modifications that had been installed in the plant during the last three years. This review included in-plant walkdowns for portions of the following installed modifications: SI and CVCS piping systems; 2A Emergency Diesel Generator (EDG)

Diagnostic/Performance Monitoring System; Unit 1 Service Water (SW) Strainer Backwash Cable Re-Route; EDG Air start system; EDG Pressure control valve setpoint modification; Unit 1 and Unit 2 SW strainers and associated Motor Operated Valve modifications. The modifications were selected based upon risk-significance, safety significance, and complexity. The inspectors reviewed the modifications selected to determine if:

• the supporting design and licensing basis documentation was updated;

• the changes were in accordance with the specified design requirements;

• the procedures and training plans affected by the modification have been adequately updated;

• the test documentation as required by the applicable test programs has been updated; and

• post-modification testing adequately verified system operability and/or functionality.

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 constituted 11 permanent plant modification samples as defined in IP 71111.17-04.

b. Findings

(1) Embedment Plate Design Deficiencies

Introduction:

The inspectors identified a finding of very low safety significance (Green)and associated Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for the failure to properly evaluate the structural steel embedment plate which supports SI pipe supports 1SI06025V and 1SI06030S. Specifically, the licensee failed to demonstrate compliance with the AISC and Seismic Category I linear elastic requirements.

Description:

The SI system is part of the Emergency Core Cooling System (ECCS).

The Braidwood Updated Final Safety Analysis Report (UFSAR), Section 6.3.1, states 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 SI system is classified as a safety Category I system in UFSAR Section 3.2.

Piping Subsystem 1SI06 is part of the SI System and is a safety-related, ASME Class II, Seismic Category I subsystem located in the curved wall area of the Auxiliary Building.

The structural steel embedment plate supports safety-related pipe supports 1SI06025V and 1SI06030S and is located in the Auxiliary Building, which is a Seismic Category I structure. The UFSAR Section 3.8.4.5.2 provides requirements for structural steel design inside the auxiliary building. Section 3.8.4.5.2 states, The stresses and strains of structural steel are limited to those specified in the AISC Specification.... Also, this section requires that stresses are held within the elastic range and no plastic deformation is allowed.

The inspectors reviewed Calculation No. 13.2.29, Structural Calculation for Mechanical Component Support 1SI06030S, Revision 4. The purpose of this calculation was to evaluate pipe support 1SI06025V and 1SI06030S structural elements for design and licensing basis requirements. The structural steel embedment plate evaluation was also contained in this calculation. The applied bending stress onto the embedment plate was greater than the allowable bending stress by 53 percent. The calculation used the following engineering judgment to justify compliance with their design and licensing basis requirements. The calculation used actual material yield stress of the embedment plate member and not specified material yield stress to calculate allowable bending stress. Also, the calculation used as an acceptance criteria, which allowed for the plastic or permanent deformation through yielding of the structural steel embedment plate and redistribution of stresses in the plate due to applied loads.

The inspectors determined that the engineering judgment used was not valid because the licensee used the actual material yield stress of material to determine the allowable bending stress as opposed to the requirement in the AISC for the allowable bending stress to use the specified minimum yield stress of the material. In addition, UFSAR Section 3.8.4.5.2 requires that no plastic or permanent deformation occur due to applied stresses. The inspectors also identified that structural steel embedment plate design loads were not correct and were non-conservative.

This issue was entered into the licensee's corrective action process as Action Request (AR) 1267356, NRC Mod/50.59 Inspection-Pipe Support Calculations, dated September 23, 2011. The licensee performed an analysis that determined the embedment plate would not experience ultimate structure failure or collapse when subjected to the design loads and determined the plate was operable but nonconforming.

Analysis:

The inspectors determined that the inadequately designed structural steel embedment plate was a performance deficiency because the structural steel embedment plate was not in conformance with AISC and Seismic Category I linear elastic requirements.

The finding was determined to be more than minor in accordance with IMC 0612 because the finding was associated with the Mitigating Systems cornerstone attribute of design control and affected the cornerstone objective of the availability, reliability, and capability of safety injection piping and pipe supports. Specifically, the licensee used the actual material yield stress to ensure the structural steel embedment plate would maintain structural integrity when subjected to design loads. This is contrary to the AISC and Seismic Category I linear elastic requirements to use the specified minimum yield stress of the material.

The inspectors determined the finding could be evaluated using the SDP in accordance with IMC 0609, Significance Determination Process, Attachment 0609.04, Phase I -

Initial Screening and Characterization of findings, Table 4a for the Mitigating Systems cornerstone. The inspectors answered yes to Question 1 under the Mitigating Systems cornerstone column of IMC 0609, Attachment 4, Table 4a, Phase I worksheet.

Specifically, the design deficiency was confirmed not to result in a loss of operability of the structural steel embedment plate. The inspectors agreed with the licensees position that the structural steel embedment plate was operable because the licensee performed an analysis that determined the embedment plate would not experience ultimate structure failure or collapse when subjected to the design loads. Therefore, the inspectors concluded that the finding did not represent an actual loss of safety function, and the issue screened out as having very low safety significance (Green).

The inspectors did not identify a cross-cutting aspect associated with this finding because the calculation was from the 1980s and was not representative of current performance.

Enforcement:

Title 10 CFR Part 50, Appendix B, Criterion III, Design Control, requires, in part, that design control measures shall provide for verifying or checking the adequacy of design, such as by the performance of design reviews, by the use of alternate or simplified calculational methods, or by the performance of a suitable testing program.

Contrary to the above, on September 23, 2011, the licensee failed to demonstrate the design adequacy of the embedment plate. Specifically, the performance of design reviews for the structural steel embedment plate were inadequate, in that Calculation No. 13.1.29 did not demonstrate that the embedment plate would meet AISC and Seismic Category I linear elastic requirements as required by the Braidwood UFSAR Section 3.8.4.5.2.

Because this violation was of very low safety significance and it was entered into the licensees corrective action program as AR 1267356, this violation is being treated as an NCV, consistent with Section 2.3.2 of the NRC Enforcement Policy. (NCV 05000456/2011008-01: Embedment Plate Design Deficiencies).

(2) Permanent Lead Shielding added to Safety Injection System and Chemical Volume and Control System Piping

Introduction:

The inspectors identified a finding of very low safety significance (Green)and associated Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for the failure to properly evaluate the Unit 1 SI subsystem 1SI06 and the Unit 1 CVCS subsystem 1CV18 piping and pipe supports. Specifically, the licensee failed to demonstrate compliance with the AISC and the ASME Boiler and Pressure Vessel Code for the 1SI06 and 1CV18 piping and pipe supports.

Description:

Braidwood UFSAR, Section 9.3.4, states the CVCS system provides safety-related seal water injection to the reactor coolant pumps and maintains required water inventory in the reactor coolant system. The CVCS system also provides control of reactor coolant water chemistry conditions, activity level, and chemical neutron absorber concentration and makeup. The CVCS system is classified as a safety Category I system in UFSAR Section 3.2.

The SI system is part of the ECCS. The Braidwood UFSAR, Section 6.3.1, states 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 is classified as a safety category I system in UFSAR Section 3.2.

The SI and CVCS piping were designed to the ASME Boiler and Pressure Vessel Code Section III and the SI and CV pipe supports were designed to the AISC code as required in UFSAR Section 3.9.3.

The inspectors reviewed Analysis No. 065613, Stress Report for Chemical Volume and Control Piping Subsystem 1CV18, Minor Revision 006M and Analysis No. 065643, Piping Stress Report for Safety Injection/Residual Heat Removal Subsystem 1SI06/1RH06, Minor Revision 004F. The calculation used NCIG 05, Guidelines for Piping System Reconciliation, Revision 1 to evaluate and accept a permanent addition of lead shielding to the 1SI06 and 1CV18 piping system.

The inspectors determined that use of NCIG-05 was not valid because the calculation did not demonstrate compliance with the AISC and ASME Section III requirements for piping and pipe supports with the addition of permanent lead shielding. Specifically, the licensee did not perform an analysis to demonstrate compliance with the AISC and ASME Section III requirements for piping and pipe supports with the addition of permanent lead shielding as required by Braidwood UFSAR Section 3.9.3.

This issue was entered into the licensee's corrective action process as AR 1269227, NRC Mod/50.59 Inspection-Use of NCIG-05 for Lead Shielding, dated September 28, 2011. The licensee performed an evaluation to demonstrate compliance with ASME Section III Appendix F operability criteria for piping and pipe supports and determined the 1SI06 and 1CV18 piping and pipe supports were operable but nonconforming.

Analysis:

The inspectors determined that the inadequately designed piping and pipe supports was a performance deficiency because the piping and pipe supports were not in conformance with AISC and ASME Boiler and Pressure Vessel Code Section III requirements with the addition of permanent lead shielding.

The finding was determined to be more than minor in accordance with IMC 0612 because the finding was associated with the Mitigating Systems cornerstone attribute of design control and affected the cornerstone objective of the availability, reliability, and capability of safety injection piping and pipe support and chemical volume and control piping and pipe supports. Specifically, the licensee did not ensure compliance with AISC and ASME Boiler and Pressure Vessel Code Section III requirements to ensure the 1SI06 and 1CV18 piping and pipe supports would maintain structural integrity when subjected to design basis loads.

The inspectors determined the finding could be evaluated using the SDP in accordance with IMC 0609, Significance Determination Process, Attachment 0609.04, Phase I -

Initial Screening and Characterization of findings, Table 4a for the Mitigating Systems cornerstone. The inspectors answered yes to Question 1 under the Mitigating Systems cornerstone column of IMC 0609, Attachment 4, Table 4a, Phase I worksheet.

Specifically, the design deficiency was confirmed not to result in a loss of operability of the 1SI06 and 1CV18 piping and pipe supports. The inspectors agreed with the licensees position that the 1SI06 and 1CV18 piping and pipe supports were operable because the licensee demonstrated compliance with ASME Section III Appendix F operability criteria for piping and pipe supports when subjected to the design loads.

Therefore, the inspectors concluded that the finding did not represent an actual loss of safety function, and the issue screened out as having very low safety significance (Green).

The inspectors did not identify a cross-cutting aspect associated with this finding because the calculational deficiency did not occur with the last three years and was not representative of current performance.

Enforcement:

Title 10 CFR Part 50, Appendix B, Criterion III, Design Control, requires, in part, that design control measures shall provide for verifying or checking the adequacy of design, such as by the performance of design reviews, by the use of alternate or simplified calculational methods, or by the performance of a suitable testing program.

Contrary to the above, on September 28, 2011, the licensee failed to demonstrate the design adequacy of 1SI06 and 1CV18 piping and pipe supports. Specifically, the performance of design reviews for 1SI06 and 1CV18 piping and pipe supports were inadequate, in that Analysis No. 065613, Minor Revision 006M and Analysis No.

065643, Minor Revision 004F did not demonstrate that the piping and pipe supports would meet ASME Section III and AISC requirements as required by the Braidwood UFSAR Section 3.9.3.

Because this violation was of very low safety significance and it was entered into the licensees corrective action program as AR 1269227, this violation is being treated as an NCV, consistent with Section 2.3.2 of the NRC Enforcement Policy. (NRC 05000456/2011008-02: Permanent Lead Shielding added to Safety Injection and Chemical Volume and Control System Piping).

4OA2 Identification and Resolution of Problems

.1 Routine Review of Condition Reports

a. Inspection Scope

From September 12 through September 30, 2011, 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 pant 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 Summary

On September 30, 2011, the inspectors presented the inspection results to Ms. A. Ferko and other members of the licensee staff. The licensee personnel acknowledged the inspection results presented and did not identify any proprietary content. The inspectors confirmed that all proprietary material reviewed during the inspection was returned to the licensee staff.

ATTACHMENT:

SUPPLEMENTAL INFORMATION

KEY POINTS OF CONTACT

Licensee

A. Ferko, Engineering Director

G. Dudek, Training Director

R. Radulovich, Nuclear Oversight Manager

P. Raush, Senior Design Engineering Manager

C. VanDenburg, Regulatory Assurance Manager

C. Mokijewski, Design Engineering

M. Grachowski, Regulatory Assurance

Nuclear Regulatory Commission

R. Daley, Chief, Engineering Branch 3, Division of Reactor Safety

J. Benjamin, Senior Resident Inspector

A. Garmoe, Reactor Inspector

LIST OF ITEMS

Opened

05000456/2011008-01; NCV Embedment Plate Design Deficiencies (Section 1R17.2.b.(1))

05000456/2011008-02; NCV Permanent Lead Shielding added to Safety Injection and Chemical Volume and Control System Piping. (Section 1R17.2.b.(2))

Closed

05000456/2011008-01; NCV Embedment Plate Design Deficiencies. (Section 1R17.2.b.(1))

05000456/2011008-02; NCV Permanent Lead Shielding added to Safety Injection and Chemical Volume and Control System Piping. (Section 1R17.2.b.(2))

Discussed

None Attachment