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| issue date = 10/28/2011
| issue date = 10/28/2011
| title = 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
| title = 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
| author name = Daley R C
| author name = Daley R
| author affiliation = NRC/RGN-III/DRS/EB3
| author affiliation = NRC/RGN-III/DRS/EB3
| addressee name = Pacilio M J
| addressee name = Pacilio M
| addressee affiliation = Exelon Generation Co, LLC, Exelon Nuclear
| addressee affiliation = Exelon Generation Co, LLC, Exelon Nuclear
| docket = 05000456, 05000457
| docket = 05000456, 05000457
Line 18: Line 18:


=Text=
=Text=
{{#Wiki_filter:UNITED STATES NUCLEAR REGULATORY COMMISSION REGION III 2443 WARRENVILLE ROAD, SUITE 210 LISLE, IL 60532
{{#Wiki_filter:October 28, 2011
-4352 October 28, 2011 Mr. Michael Senior Vice President, Exelon Generation Company, LLC President and Chief Nuclear Officer (CNO), Exelon Nuclear 4300 Winfield Road Warrenville, IL 60555


SUBJECT: BRAIDWOOD STATION, UNIT S 1 AND 2 EVALUATION OF CHANGES, TESTS, OR EXPERIMENTS AND PERMANENT PLANT MODIFICATIONS BASELINE INSPECTION REPORT 05000456/2011008; 05000457/2011008 (DRS)
==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:==
==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.
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 Commission's rules and regulations and with the conditions of your license. The inspectors reviewed selected procedures and records, observed activities, and interviewed personnel.
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.


Based on the results of this inspection, two NRC
The inspectors reviewed selected procedures and records, observed activities, and interviewed personnel.
-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
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.
-0001; wit h 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.
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.


50-456; 50-457 License Nos.
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).


NPF-72; NPF-77  
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:===
===Enclosure:===
Inspection Report 05000456/2011008; 05000457/2011008;
Inspection Report 05000456/2011008; 05000457/2011008; w/Attachment: Supplemental Information


===w/Attachment:===
REGION III==
Supplemental Information cc w/encl: Distribution via ListServ
Docket No:
 
Enclosure U. S. NUCLEAR REGULATORY COMMISSION REGION III Docket No:
50-456; 50-457 License No:
50-456; 50-457 License No:
NPF-72; NPF-77 Report 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
05000456/2011008; 05000457/2011008 Licensee:
- 30, 2011 Inspectors:
Exelon Generation Company, LLC Facility:
J. Bozga, Reactor Inspector (Lead)
Braidwood Station, Units 1 and 2 Location:
J. Gilliam, Reactor Inspector M. Jones, Reactor Inspector Approved by:
Braceville, IL Dates:
R. Daley, Chief Engineering Branch 3 Division of Reactor Safety 1 Enclosure  
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=
=SUMMARY OF FINDINGS=
IR 05000456/2011008, 05000457/2011008; 09/12/2011  
IR 05000456/2011008, 05000457/2011008; 09/12/2011 - 09/30/2011; Braidwood Station


- 09/30/2011; Braidwood Station Units 1 and 2; Evaluation of Changes, Tests, or Experiments and Permanent Plant Modifications.
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.
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 inspection was conducted by Region III based engineering inspectors. Two NRC
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,
-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 NRC's program for overseeing the safe operation of commercial nuclear power reactors is described in NUREG
Revision 4, dated December 2006.
-1649, "Reactor Oversight Process," Revision 4, dated December 2006.


A.
A.


===Cornerstone: Mitigating Systems===
===Cornerstone: Mitigating Systems===
===NRC-Identified===
===NRC-Identified===
and Self-Revealed Findings GreenThe 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.
and Self-Revealed Findings
* Green 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))
: 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 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))
*
:  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.
: '''Green.'''
: '''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 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 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.
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))
B.


(Section 1R17.
No violations of significance were identified.
 
2.b.(2)) B. No violations of significance were identified.


===Licensee-Identified Violations===
===Licensee-Identified Violations===
1.


1. REACTOR SAFETY
REACTOR SAFETY


=REPORT DETAILS=
=REPORT DETAILS=


===Cornerstone:===
===Cornerstone: Initiating Events, Mitigating Systems, and Barrier Integrity===
Initiating Events, Mitigating Systems, and Barrier Integrity
 
{{a|1R17}}
{{a|1R17}}
==1R17 Evaluation==


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


===.1 ===
==
===.1===
{{IP sample|IP=IP 71111.17}}
{{IP sample|IP=IP 71111.17}}
a. Evaluation of Changes, Tests, or Experiments 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:
a.
Inspection Scope 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.
 
Evaluation of Changes, Tests, or Experiments 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:
Inspection Scope
* 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
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.
-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. No findings of significance were identifiedFindings 4 Enclosure
This inspection constituted six samples of evaluations and 15 samples of changes as defined in IP 71111.17-04.


===.2 a. Permanent Plant Modifications===
b.


From September 12, 2011 through September 30, 20 11, the inspectors reviewed 11 permanent plant modifications that had been installed in the plant during the last three years. This review included in
No findings of significance were identified Findings
-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
===.2 a.===
-Route; EDG Air start system
Permanent Plant Modifications 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)
; 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
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:
-significance, safety significance, and complexity. The inspectors reviewed the modifications selected to determine if:
Inspection Scope
Inspection Scope 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 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.
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. (1) Findings Embedment Plate Design Deficiencies  
This inspection constituted 11 permanent plant modification samples as defined in IP 71111.17-04.
 
b.
: (1) Findings Embedment Plate Design Deficiencies  


=====Introduction:=====
=====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.
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
=====Description:=====
:  The SI system is part of the Emergency Core Cooling System (ECCS).
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 5 Enclosure 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.
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
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.
-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
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.
-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 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 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
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.
, 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
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.
, 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
Analysis 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  
-Pipe Support Calculations
: 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.
," 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.


AnalysisThe 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 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.


6 Enclosure maintain structural integrity when subjected to design 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.
. 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 licensee's 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).
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
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.
-cutting aspect associated with this finding because the calculation was from the 1980s and was not representative of current performance.


EnforcementContrary 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.
Enforcement 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.
: 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.


:  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.
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


Because this violation was of very low safety significance and it was entered into the licensee's 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
=====Introduction:=====
-01:  Embedment Plate Design Deficiencies).
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.


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


=====Introduction:=====
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 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.
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.


Description
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.
:  Braidwood UFSAR , Section 9.3.4, states the CVCS system provide s 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 7 Enclosure 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 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.


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.
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.


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.
Analysis 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 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 inspectors determined that use of NCIG
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.
-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
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.
-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.


AnalysisThe 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.
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 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 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.


The inspectors determined the finding could be evaluated using the SDP in accordance with IMC 0609, "Significance Determination Process," Attachment 0609.04, "Phase I
Enforcement 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.
-
8 Enclosure 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 licensee's 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.


EnforcementContrary 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.
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.
: 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.


:  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.
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).


Because this violation was of very low safety significance and it was entered into the licensee's 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
==OTHER ACTIVITIES (OA)==
-02:  Permanent Lead Shielding added to Safety Injection and Chemical Volume and Control System Piping).
4OA2


==OTHER ACTIVITIES==
(OA
)
{{a|4OA2}}
==4OA2 ==
===.1 Identification and Resolution of Problems===
===.1 Identification and Resolution of Problems===
a.


a. Routine Review of Condition Reports 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. Inspection Scope
Routine Review of Condition Reports 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.


9 Enclosure b. No finding s of significance were identified.
Inspection Scope b.


Findings
No findings of significance were identified.
{{a|4OA6}}
 
==4OA6 ==
Findings 4OA6
===.1 Meetings 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.
===.1 Meetings===
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.


===Exit Meeting Summary===
===Exit Meeting Summary===
 
ATTACHMENT:  
ATTACHMENT:


=SUPPLEMENTAL INFORMATION=
=SUPPLEMENTAL INFORMATION=


Attachment
==KEY POINTS OF CONTACT==
SUPPLEMENTAL INFORMATION KEY POINTS OF CONTAC
T
: [[contact::A. Ferko]], Engineering Director
: [[contact::A. Ferko]], Engineering Director
Licensee  
Licensee  
: [[contact::G. Dudek]], Training Director
: [[contact::G. Dudek]], Training Director  
: [[contact::R. Radulovich]], Nuclear Oversight Manager  
: [[contact::R. Radulovich]], Nuclear Oversight Manager  
: [[contact::P. Raush]], Senior Design Engineering Manager
: [[contact::P. Raush]], Senior Design Engineering Manager  
: [[contact::C. VanDenburg]], Regulatory Assurance Manager
: [[contact::C. VanDenburg]], Regulatory Assurance Manager  
: [[contact::C. Mokijewski]], Design Engineering
: [[contact::C. Mokijewski]], Design Engineering  
: [[contact::M. Grachowski]], Regulatory Assurance
: [[contact::M. Grachowski]], Regulatory Assurance  
: [[contact::R. Daley]], Chief, Engineering Branch 3, Division of Reactor Safety
: [[contact::R. Daley]], Chief, Engineering Branch 3, Division of Reactor Safety
Nuclear Regulatory Commission
Nuclear Regulatory Commission  
: [[contact::J. Benjamin]], Senior Resident Inspector
: [[contact::J. Benjamin]], Senior Resident Inspector  
: [[contact::A. Garmoe]], Reactor Inspector
: [[contact::A. Garmoe]], Reactor Inspector  
LIST OF ITEMS OPENED, CLOSED AND DISCUSS
 
ED 05000456/2011008
==LIST OF ITEMS==
-01; Opened NCV Embedment Plate Design Deficiencies (Section 1R17.
===OPENED, CLOSED AND DISCUSSED===
2.b.(1)) 05000456/2011008
: 05000456/2011008-01;  
-02;  NCV Permanent Lead Shielding added to Safety Injection and Chemical Volume and Control System Piping.  (Section 1R17.2.b.(2)) 05000456/2011008
 
-01;  Closed NCV Embedment Plate Design Deficiencies.  (Section 1R17.2.b.(1)) 05000456/2011008
===Opened===
-02; NCV Permanent Lead Shielding added to Safety Injection and Chemical Volume and Control System
NCV Embedment Plate Design Deficiencies (Section 1R17.2.b.(1))  
Piping.  (Section 1R17.2.b.(2)) NoneDiscussed
: 05000456/2011008-02;  
Attachment
LIST OF DOCUMENTS REVIEWED The following is a list of documents reviewed during the inspection. Inclusion on this list does not imply that the NRC inspectors reviewed the documents in their entirety, but rather, that selected sections of portions of the documents were evaluated as part of the overall inspection effort. Inclusion of a document on this list does not imply NRC acceptance of the document or any part of it, unless this is stated in the body of the inspection report.
CALCULATIONS
Number Description or Title
48630 Date or Revision Piping Stress Report For Subsystem 1SI16
49258 Piping Calculation for Subsystem 1CV21, Chemical and Volume Control
001E 53790 Addendum to Piping Stress Report Chemical & Volume Control System 2CV53
000E 65613 Stress Report for Chemical
and Volume Control Piping System 1CV1
006M 65643 Piping Stress Report for Safety Injection/Residual Heat Removal Subsystem 1SI06/1RH06
004F 13.2.2-BRW-08-0049-S Qualification of Existing Pipe Support M
-2AF08005G for Revised Loads
13.2.2-BRW-08-0051-S Qualification of Existing Pipe Support M
-2AF08014R for Revised Loads
13.2.2-BRW-08-0052-S Qualification of Existing Modified Pipe Support M
-2AF08023G for Revised Loads
13.1.2-BRW-08-0053-S Qualification of Existing Pipe Support M
-1AF08024G for Revis
ed Loads 0 13.2.18BR-BRW-03-0015-S Mechanical Component Supports 0IA74A2S184T
and 0IA74A2S185T
13.2.29BR Mechanical Component Support 1SI06030S
BRW-00-0010-M Byron/Braidwood Uprate Project
- Spent Fuel Pool Temperature Analysis
BRW-08-0074 Auxiliary Feedwater Cross Tie Hydraulic Analysis
BRW-98-0724-E Motor Operated Valve (MOV)
Actuator Terminal Voltage and Thermal Overload Sizing Calculation
-SX System 0 BRW-96-089-M Verification of Braidwood 125 VDC Battery 111, 112, 211 and 212 Ventilation Requirements and Hydrogen Concentration Evaluation Following a
Loss of Battery
Room 1 SITH-1 Refueling Water Storage Tank (RWST) Level Setpoints 7
Attachment
CORRECTIVE ACTION PROGRAM DOCUMENTS REVIEWED
Number Description or Title
0562372 Date or Revision Byron CDBI Issue
- CALC BRW-04-0005-M/BYR04-016 November 27, 2006 0813142 Abnormal System Response During 1SX01PA ASME September 2 , 2008 0882419 1B DG 30# Regulator Needs Adjusted
- 1DG5228B February 18, 2009
0908920 2A DG Recorder Reading High, Verify Calibration
April 18, 2009
0914792 Determination of Sx Strainer Manipulation on the Sx System May 1, 2009
0960770 Overhaul for MOV 1SX150B and Diagnostically Test Sept 2, 2009
0960766 Overhaul Actuator for MOV 1SX150A and Diagnostically Test
October 2, 2009
22480 2SX150A SX Late Changes In the On
-Line Work Scheduling
- 2SX150A February 27, 2010
1098702 EC 374543 Late Issuance Challenges Station
Resources August 6, 2010
1117907 U-2 RCS Arrow Range RTD Drifting
Sept 26, 2010 1128271  Sx MOD Cost Impacted by Missed Engineering Milestones
October 19, 2010
1132815 Wiring in MCC 133X1
-A-D1 and D4 cannot be Restored October 29, 2010
1171787 1A DG 40 No. Regulator Requires Replacement
- 1DG5228A February 7, 2011
CORRECTIVE ACTION PROGRAM DOCUMENTS GENERATED
Number Description or Title
267356 Date or Revision
NRC Mod/50.59 Inspection
- Pipe Support Calculations
Sept 23, 2011 1268631 TRM Implementation not Timely
Sept 27, 2011
269263 NRC Mod/50.59 Inspection
- Concrete Compressive Strength
Sept 27, 2011
269227 NRC Mod 50.59 Inspection
- Use of NCIG
-05 for Lead Shielding
Sept 28, 2011


Attachment
NCV Permanent Lead Shielding added to Safety Injection and Chemical Volume and Control System Piping. (Section 1R17.2.b.(2))
DRAWINGS Number Description or Title
: 05000456/2011008-01;
1A-AF-34 Date or Revision
Auxiliary Feedwater Large Bore Isometric
B 2A-AF-29 Auxiliary Feedwater Auxiliary Building
E 20E-1-4030DG32 Schematic Diagram Diesel Generator 1A Starting Sequence Control 1 DG01KA Part 2
AI M-1SI06025V Support No. M
-1SI06025V H M-1SI06030S Support No. M
-1SI06030S G M-37 Diagram of Auxiliary Feedwater Unit 1
BG M-122 Diagram of Auxiliary Feedwater Unit 2
BA M-152 Manufacturer's Supplemental Diagram of Diesel Generator Control Diagram Shutdown System Units 1
and 2 L S-497 Turbine Room Concrete Column Schedule
AC S-1404 Refueling Water Storage
Tank Section and Details W  10 CFR 50.59 EVALUATIONS
Number Description or Title
BRW-E-2008-154 Date or Revision
Zinc Injection into the CV Letdown Flow Path
Sept, 25, 2008
BRW-E-2008-168 Modify U2 RWST Level Transmitter Drain Line 2SI99F
-1" to Eliminate Single Point of Vulnerability
October 8, 2010
BRW-E-2010-10 Zinc Injection into the Unit 1 CV Letdown Flow Path May 14, 2010
BRW-E-2010-57 Rewire/Bypass Sensors
and 4 in Order to Restore Remaining 2B RVLIS Probe Sensors
June 28, 2010
BRW-E-2010-126 Change In-Core Decay Time
for A1R15
October 2, 2010
BRW-E-2010-163 2C DT/TA Loop
-Two RTD Operation for Cycle BR2C15 October 30, 2010
CFR 50.59 SCREENINGS
Number Description or Title
BRW-S-2008-150 Date or Revision
ECs to Provide Temporary Reliable Power Feed to SX Strainers
Sept 4, 2008
BRW-S-2008-151 Revised Procedure for Installation of Safety-Related Power Supplies to Sx Strainers. ECs 372033, 372036, 372037, 372044 October 9, 2008
BRW-S-2009-4 Installation of Polymer Dispersant Injection System to Unit 2 Feedwater System
January 13, 2009
BRW-S-2009-15 Revised NPSH Values for the CV and SI April 3, 2009


Attachment
===Closed===
CFR 50.59 SCREENINGS
NCV Embedment Plate Design Deficiencies. (Section 1R17.2.b.(1))  
Number Description or Title
: 05000456/2011008-02; NCV Permanent Lead Shielding added to Safety Injection and Chemical Volume and Control System Piping. (Section 1R17.2.b.(2))
Pumps Date or Revision
None
BRW-S-2009-23 Replace Diesel Generator Starting Air Dryer 2DG01SB
-D with New Model
February 13, 2009
BRW-S-2009-92 U-1 (U-2) Rod Drive Power Supply
-Addition of Load Resisters
Sept 28, 2009
BRW-S-2009-124 2SI06-Modify Supports due to
Re-Analysis per NRC finding
October 7, 2009
BRW-S-2009-135 Emergency Diesel Generator Governor Booster Modification
December 4, 2009
BRW-S-2010-164 Develop Calculation Supporting AF Diesel Fuel Storage Tech Spec Requirement and
revise Tech Spec Basis
December 21, 2010 BRW-S-2010-172 Temporary Scaffold for WO No. 00612535 per AR 01130014
-03 Nov 12, 2010
BRW-S-2010-177 Temporary Scaffold for WO #1236986
Nov 23, 2010
BRW-S-2011-35 Disable Alarms for Fire Detection Zone
2D-31 (2W MPT)
Sept 7, 2011
BRW-S-2011-54 Containment Fuel Transfer Area Permanent Supports and Removable Screen Doors
April 7, 2011
BRW-S-2011-75 Disable Iso Phase Bus Duct Temp Switch ITS-MP031 May 11, 2011
BRW-S-2011-84 Temporary Scaffold for WO No. 01275408 and WO No. 01274736 June 9, 2011
MODIFICATIONS
Number Description or Title
41716 Date or Revision
Add Supports for Air Line to 2SX114B, Containment Chiller
360532 Change Setpoint from 30 PSIG to 40 PSIG for the 1DG5228A/B and 1DG5231A/B
Pressure Control Valves
364360 U2 Permanent Lead Shielding  
- Exelon TSP's
01-043 to 045, 049
-051, 053, 03
-033, -035, 036, and 052. Systems CV, RH, SI
366445 U0 and U1 Permanent Lead Shielding
- S and L TSP's 01
-023, 040, 052, 02
-037, 039, 03
-009, 010, 057, 058, 06
-026, 041, 0
0 369972 AFW Cross Tie
370521 2A Diesel Generator Diagnostic Performance Monitoring System
Attachment
MODIFICATIONS
Number Description or Title
2033 Date or Revision
Provide Temporary Reliable Power Feed to 1A SX Strainer
373770 Unit 1 SX Strainer Upgrade
374543 Unit 1 SX Strainer Backwash Cable Re
-Route 4 376814 Revise Calc BRW
-96-089 to Evaluate Battery Room Ventilation Requirements for the Existing 125 VDC Batteries
381707 Re-Analysis of Piping Subsystem 1SI16
OTHER DOCUMENTS
Number Description or Title
1Bw0A Date or Revision
Essential Service Water Malfunction Unit 1
Revision 102
ATI 813142
Bryzoa Deposition And Growth in the CW Forebays Resulted in Rapid Fouling Of SX Strainers, and Inoperability of the 1A and 2A SX Trains December 10, 2008 BwMP SX Strainer Manual Backwash Operation of Loss of Power Revision 1
BwOP AF-3 Filling and Venting the Auxiliary Feedwater
System Revision 26
BwOP SX-6 Essential Service Water Strainer Manual Operation Revision 8
CC-AA-112 Temporary Configuration Changes
Revision 17
ER-AA-300 Motor Operated Valve Program Administrative Procedure Revision 6
FDRP 24-004 Revised Fire Loading in the Unit 1 Turbine Building Mezzanine Floor
October 12, 2008
LA-AA-107-1001 UFSAR Update T&RM
Revision 1
LS-AA-104 Exelon 50.59 Review Process
Revision 6
TRM 11-003 Revise Technical Requirements Manual (TRM) Table T3.8.b
-1, "Thermal Overload Protection Device-Unit 1", Adding Valves 1SX150A/B
June 9, 2011
TRP 1PI-DG8040A Calibration of DG 1A Control Air Valve 1 DG 5231A Pressure Indicator
November 18, 2010 WO 0126662
Overhaul Actuator for MOV 2SX150A and Diagnostically Test
February 17, 2011


Attachment
===Discussed===
LIST OF ACRONYMS USE
==LIST OF DOCUMENTS REVIEWED==
D ADAMS Agencywide Documents Access and Management System
AISC  American Institute of Steel Construction
AR  Action Request
ASME  American Society of Mechanical Engineers
CFR  Code of Federal Regulations
CNO  Chief Nuclear Officer
CVCS  Chemical Volume and Control System
DRS  Division of Reactor Safety
EC  Engineering Change
ECCS  Emergency Core Cooling System
EDG  Emergency Diesel Generator
IMC  Inspection Manual Chapter
NCV  Non-Cited Violation
NRC  U.S. Nuclear Regulatory Commission
PARS  Public Available Records System
SDP  Significance Determination Process
SI  Safety Injection
SW  Service Water
UFSAR Updated Final Safety Analysis Report
Attachment
M. Pacilio
    -2- 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
: [[contact::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
cc w/encl:
Distribution via ListServ
DISTRIBUTIONSee next page
: DOCUMENT NAME:
G:\DRSIII\DRS\Work in Progress
\BRDWD 2011
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Attachment
Letter to Mr. Mich
ael J. Pacilio from Mr. Robert C. Daley dated October
8, 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
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Latest revision as of 22:46, 12 January 2025

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  Constellation icon.png
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)
v • d • e


Text

October 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.

A.

Cornerstone: Mitigating Systems

NRC-Identified

and Self-Revealed Findings

  • Green 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))
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.
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))

B.

No violations of significance were identified.

Licensee-Identified Violations

1.

REACTOR SAFETY

REPORT DETAILS

Cornerstone: Initiating Events, Mitigating Systems, and Barrier Integrity

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

==

.1

Main article: IP 71111.17

a.

Evaluation of Changes, Tests, or Experiments 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:

Inspection Scope

  • 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.

No findings of significance were identified Findings

.2 a.

Permanent Plant Modifications 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:

Inspection Scope

  • 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.

(1) Findings 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 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

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.

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 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.

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.

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 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 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 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 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.

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.

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).

OTHER ACTIVITIES (OA)

4OA2

.1 Identification and Resolution of Problems

a.

Routine Review of Condition Reports 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.

Inspection Scope b.

No findings of significance were identified.

Findings 4OA6

.1 Meetings

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.

Exit Meeting Summary

ATTACHMENT:

SUPPLEMENTAL INFORMATION

KEY POINTS OF CONTACT

A. Ferko, Engineering Director

Licensee

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

Nuclear Regulatory Commission

J. Benjamin, Senior Resident Inspector
A. Garmoe, Reactor Inspector

LIST OF ITEMS

OPENED, CLOSED AND DISCUSSED

05000456/2011008-01;

Opened

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))

05000456/2011008-01;

Closed

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))

None

Discussed

LIST OF DOCUMENTS REVIEWED

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