IR 05000282/2006006: Difference between revisions

From kanterella
Jump to navigation Jump to search
(Created page by program invented by StriderTol)
(Created page by program invented by StriderTol)
 
(2 intermediate revisions by the same user not shown)
Line 3: Line 3:
| issue date = 04/28/2006
| issue date = 04/28/2006
| title = IR 05000282-06-006 (Drs); 05000306-06-006 (Drs); 03/06/2006 - 03/24/2006; Prairie Island Nuclear Generating Plant, Units 1 and 2; Evaluation of Changes, Tests, or Experiments (10 CFR 50.59) and Permanent Plant Modifications
| title = IR 05000282-06-006 (Drs); 05000306-06-006 (Drs); 03/06/2006 - 03/24/2006; Prairie Island Nuclear Generating Plant, Units 1 and 2; Evaluation of Changes, Tests, or Experiments (10 CFR 50.59) and Permanent Plant Modifications
| author name = Hills D E
| author name = Hills D
| author affiliation = NRC/RGN-III/DRS/EB1
| author affiliation = NRC/RGN-III/DRS/EB1
| addressee name = Palmisano T
| addressee name = Palmisano T
Line 18: Line 18:


=Text=
=Text=
{{#Wiki_filter:
{{#Wiki_filter:==SUBJECT:==
[[Issue date::April 28, 2006]]
PRAIRIE ISLAND NUCLEAR GENERATING PLANT, UNITS 1 AND 2 NRC EVALUATION OF CHANGES, TESTS, OR EXPERIMENTS AND PERMANENT PLANT MODIFICATIONS BASELINE INSPECTION REPORT 05000282/2006006 (DRS); 05000306/2006006 (DRS)


SUBJECT: PRAIRIE ISLAND NUCLEAR GENERATING PLANT, UNITS 1 AND 2 NRC EVALUATION OF CHANGES, TESTS, OR EXPERIMENTS ANDPERMANENT PLANT MODIFICATIONS BASELINE INSPECTION REPORT 05000282/2006006 (DRS); 05000306/2006006 (DRS)
==Dear Mr. Palmisano:==
On March 24, 2006, the U.S. Nuclear Regulatory Commission (NRC) completed a combined baseline inspection of the Evaluation of Changes, Tests, or Experiments and Permanent Plant Modifications at the Prairie Island Nuclear Generating Plant. The enclosed report documents the results of the inspection, which were discussed and others of your staff at the completion of the inspection on March 24, 2006.
 
The inspectors examined activities conducted under your license as they relate to safety and compliance with the Commissions rules and regulations and with the conditions of your license.
 
The inspectors reviewed selected procedures and records, observed activities, and interviewed personnel.


==Dear Mr. Palmisano:==
Based on the results of the inspection, one NRC-identified finding of very low safety significance was identified which involved a violation of NRC requirements. However, because this violation was of very low safety significance, not willful, and because it was entered into your corrective action program, the NRC is treating the issue as a Non-Cited Violation in accordance with Section VI.A.1 of the NRCs Enforcement Policy.
On March 24, 2006, the U.S. Nuclear Regulatory Commission (NRC) completed a combinedbaseline inspection of the Evaluation of Changes, Tests, or Experiments and Permanent Plant Modifications at the Prairie Island Nuclear Generating Plant. The enclosed report documents the results of the inspection, which were discussed and others of your staff at thecompletion of the inspection on March 24, 2006.The inspectors examined activities conducted under your license as they relate to safety andcompliance 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.Based on the results of the inspection, one NRC-identified finding of very low safetysignificance was identified which involved a violation of NRC requirements. However, becausethis violation was of very low safety significance, not willful, and because it was entered intoyour corrective action program, the NRC is treating the issue as a Non-Cited Violation in accordance with Section VI.A.1 of the NRC's Enforcement Policy.If you contest the subject or severity of a Non-Cited Violation, you should provide a responsewithin 30 days of the date of this inspection report, with the basis for your denial, to the U.S.
If you contest the subject or severity of a Non-Cited Violation, 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-
Nuclear Regulatory Commission, ATTN: Document Control Desk, Washington, DC 20555-0001, with a copy to the Regional Administrator, U.S. Nuclear Regulatory Commission -
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 Prairie Island Nuclear Generating Plant facility.
Region III, 2443 Warrenville Road, Suite 210, Lisle, IL 60532-4352; the Director, Office ofEnforcement, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001; and the Resident Inspector Office at the Prairie Island Nuclear Generating Plant facility.


In accordance with 10 CFR 2.390 of the NRC's "Rules of Practice," a copy of this letterand its enclosure will be available electronically for public inspection in the NRC PublicDocument Room or from the Publicly Available Records (PARS) component of NRC's document system (ADAMS). ADAMS is accessible from the NRC Web site at http://www.nrc.gov/reading-rm/adams.html (the Public Electronic Reading Room).
In accordance with 10 CFR 2.390 of the NRC's "Rules of Practice," a copy of this letter and its enclosure will be available electronically for public inspection in the NRC Public Document Room or from the Publicly Available Records (PARS) component of NRC's document system (ADAMS). ADAMS is accessible from the NRC Web site at http://www.nrc.gov/reading-rm/adams.html (the Public Electronic Reading Room).


Sincerely,/RA/David E. Hills, ChiefEngineering Branch 1 Division of Reactor SafetyDocket Nos. 50-282; 50-306License Nos.
Sincerely,
/RA/
David E. Hills, Chief Engineering Branch 1 Division of Reactor Safety Docket Nos. 50-282; 50-306 License Nos.


===Enclosure:===
===Enclosure:===
Inspection Report 05000282/2006006 (DRS); 05000306/2006006 (DRS)
Inspection Report 05000282/2006006 (DRS); 05000306/2006006 (DRS)
Document Room or from the Publicly Available Records (PARS) component of NRC's document system (ADAMS). ADAMS is accessible from the NRC Web site at http://www.nrc.gov/reading-rm/adams.html (the Public Electronic Reading Room).David E. Hills, ChiefEngineering Branch 1 Division of Reactor SafetyDocket Nos. 50-282; 50-306License Nos. DPR-42; DPR-60


===Enclosure:===
Document Room or from the Publicly Available Records (PARS) component of NRC's document system (ADAMS). ADAMS is a
Inspection Report 05000282/2006006 (DRS); 05000306/2006006 (DRS)
 
cc w/encl:C. Anderson, Senior Vice President, Group OperationsM. Sellman, Chief Executive Officer and Chief Nuclear Officer Regulatory Affairs Manager J. Rogoff, Vice President, Counsel and Secretary Nuclear Asset Manager Tribal Council, Prairie Island Indian Community Administrator, Goodhue County Courthouse Commissioner, Minnesota Department of Commerce Manager, Environmental Protection Division Office of the Attorney General of MinnesotaDOCUMENT NAME:E:\Filenet\ML061220751.wpd G Publicly Available G Non-Publicly Available G Sensitive G Non-SensitiveTo receive a copy of this document, indicate in the concurrence box "C" = Copy without attach/encl "E" = Copy with attach/encl "N" = No copyOFFICERIIIRIIIRIIIRIIINAMEJNeurauterRSkokowskiDHillsDATE04/10/0604/27/0604/28/06OFFICIAL RECORD COPY T. Palmisano-3-
REGION III==
U.S. NUCLEAR REGULATORY COMMISSIONREGION IIIDocket No:50-282; 50-306License No:Report No:05000282/2006006 (DRS); 05000306/2006006 (DRS)Licensee:Facility:Prairie Island Nuclear Generating PlantLocation:Dates:March 6 through March 24, 2006 Inspectors:J. Neurauter, Senior Reactor Inspector, Team LeaderAlan Dahbur, Reactor InspectorApproved by:D. Hills, ChiefEngineering Branch 1 Division of Reactor Safety (DRS)  
Docket No: 50-282; 50-306 License No:
Report No: 05000282/2006006 (DRS); 05000306/2006006 (DRS)
Licensee:
Facility: Prairie Island Nuclear Generating Plant Location:
Dates: March 6 through March 24, 2006 Inspectors: J. Neurauter, Senior Reactor Inspector, Team Leader Alan Dahbur, Reactor Inspector Approved by: D. Hills, Chief Engineering Branch 1 Division of Reactor Safety (DRS)


=SUMMARY OF FINDINGS=
=SUMMARY OF FINDINGS=
IR 05000282/2006006 (DRS); 05000306/2006006 (DRS); 03/06/2006 - 03/24/2006; PrairieIsland Nuclear Generating Plant, Units 1 and 2;  Evaluation of Changes, Tests, or Experiments (10 CFR 50.59) and Permanent Plant Modifications.The inspection covered a two-week announced baseline inspection on evaluations of changes,tests, or experiments and permanent plant modifications. The inspection was conducted by two regional based engineering inspectors. One Green Non-Cited Violation (NCV) was identified.
IR 05000282/2006006 (DRS); 05000306/2006006 (DRS); 03/06/2006 - 03/24/2006; Prairie


The significance of most findings is indicated by their color (Green, White, Yellow, Red), using Inspection Manual Chapter 0609, "Significance Determination Process (SDP.)"  Findings for which the SDP does not apply, may be Green, or be assigned a severity level after NRC management review. The NRC's program for overseeing the safe operation of commercialnuclear power reactors is described in NUREG-1649, "Reactor Oversight Process," Revision 3,dated July 2000.A.Inspector-Identified and Self-Revealed Findings
Island Nuclear Generating Plant, Units 1 and 2; Evaluation of Changes, Tests, or Experiments (10 CFR 50.59) and Permanent Plant Modifications.
 
The inspection covered a two-week announced baseline inspection on evaluations of changes, tests, or experiments and permanent plant modifications. The inspection was conducted by two regional based engineering inspectors. One Green Non-Cited Violation (NCV) was identified.
 
The significance of most findings is indicated by their color (Green, White, Yellow, Red), using Inspection Manual Chapter 0609, Significance Determination Process (SDP.) Findings for which the SDP does not apply, may be Green, or be assigned a severity level after NRC management review. The NRC's program for overseeing the safe operation of commercial nuclear power reactors is described in NUREG-1649, Reactor Oversight Process, Revision 3, dated July 2000.
 
A.     Inspector-Identified and Self-Revealed Findings


===Cornerstone: Mitigating Systems===
===Cornerstone: Mitigating Systems===
: '''Green.'''
: '''Green.'''
A Non-Cited violation of 10 CFR Part 50, Appendix B, Criterion III, "DesignControl," having very low safety significance was identified by the inspectors.
A Non-Cited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, having very low safety significance was identified by the inspectors.


Specifically, the licensee had not evaluated and updated the associated plant cableampacity calculation to determine the potential consequences of adverse effects to cabling due to higher temperatures in the auxiliary feedwater (AFW) pump rooms andother auxiliary building areas. After identification by the inspectors, the licensee wasable to demonstrate that even though the higher temperatures decreased the ampacitymargins for the affected cabling, it did not decrease the margins to the limit where the cabling would fail if called upon to provide power to equipment important to safety.The finding was more than minor because it affected the mitigating system cornerstoneobjective to ensure the availability, reliability, and capability of systems that mitigatetransients and accidents, and if left uncorrected, the finding could become a more significant safety concern. Specifically, if left uncorrected, the licensee may not account for high temperature conditions in plant areas that could adversely affect the ampacity of cabling that supply power to equipment important to safety. This finding was of very low safety significance because, the licensee's preliminary evaluation determined that thehigher temperatures in the AFW pump rooms and other auxiliary building areas wouldnot prevent equipment important to safety from functioning. (Section 1R17.1.b.1)
Specifically, the licensee had not evaluated and updated the associated plant cable ampacity calculation to determine the potential consequences of adverse effects to cabling due to higher temperatures in the auxiliary feedwater (AFW) pump rooms and other auxiliary building areas. After identification by the inspectors, the licensee was able to demonstrate that even though the higher temperatures decreased the ampacity margins for the affected cabling, it did not decrease the margins to the limit where the cabling would fail if called upon to provide power to equipment important to safety.
 
The finding was more than minor because it affected the mitigating system cornerstone objective to ensure the availability, reliability, and capability of systems that mitigate transients and accidents, and if left uncorrected, the finding could become a more significant safety concern. Specifically, if left uncorrected, the licensee may not account for high temperature conditions in plant areas that could adversely affect the ampacity of cabling that supply power to equipment important to safety. This finding was of very low safety significance because, the licensees preliminary evaluation determined that the higher temperatures in the AFW pump rooms and other auxiliary building areas would not prevent equipment important to safety from functioning. (Section 1R17.1.b.1)


===Cornerstone: Barrier Integrity===
===Cornerstone: Barrier Integrity===
Line 60: Line 76:
No findings of significance were identified.
No findings of significance were identified.


===B.Licensee-Identified Violations===
===Licensee-Identified Violations===
 
None.
None.


=REPORT DETAILS=
=REPORT DETAILS=
1.REACTOR SAFETYCornerstones: Initiating Events, Mitigating Systems, and Barrier Integrity1R02Evaluations of Changes, Tests, or Experiments (71111.02).1Review of 10 CFR 50.59 Evaluations and Screenings
 
==REACTOR SAFETY==
Cornerstones: Initiating Events, Mitigating Systems, and Barrier Integrity {{a|1R02}}
==1R02 Evaluations of Changes, Tests, or Experiments==
{{IP sample|IP=IP 71111.02}}
===.1 Review of 10 CFR 50.59 Evaluations and Screenings===


====a. Inspection Scope====
====a. Inspection Scope====
From March 6 through March 24, 2006, the inspectors reviewed eight evaluationsperformed pursuant to 10 CFR 50.59. The inspectors confirmed that the evaluationswere thorough and that prior NRC approval was obtained as appropriate. Theinspectors also reviewed seventeen screenings where licensee personnel had determined that a 10 CFR 50.59 evaluation was not necessary. In regard to the changes reviewed where no 10 CFR 50.59 evaluation was performed, the inspectors verified that the changes did not meet the threshold to require a 10 CFR 50.59evaluation. The evaluations and screenings were chosen based on risk significance, safety significance, and complexity. The list of documents reviewed by the inspectors is included as an attachment to this report.The inspectors used, in part, Nuclear Energy Institute (NEI) 96-07, "Guidelines for 10 CFR 50.59 Implementation," Revision 1, to determine acceptability of the completed evaluations and screenings. The NEI document was endorsed by the NRC inRegulatory 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."
From March 6 through March 24, 2006, the inspectors reviewed eight evaluations performed pursuant to 10 CFR 50.59. The inspectors confirmed that the evaluations were thorough and that prior NRC approval was obtained as appropriate. The inspectors also reviewed seventeen screenings where licensee personnel had determined that a 10 CFR 50.59 evaluation was not necessary. In regard to the changes reviewed where no 10 CFR 50.59 evaluation was performed, the inspectors verified that the changes did not meet the threshold to require a 10 CFR 50.59 evaluation. The evaluations and screenings were chosen based on risk significance, safety significance, and complexity. The list of documents reviewed by the inspectors is included as an attachment to this report.
 
The inspectors used, in part, Nuclear Energy Institute (NEI) 96-07, Guidelines for 10 CFR 50.59 Implementation, Revision 1, to determine acceptability of the completed evaluations and screenings. The NEI document was endorsed by the NRC in Regulatory Guide 1.187, Guidance for Implementation of 10 CFR 50.59, Changes, Tests, and Experiments, dated November 2000. The inspectors also consulted Part 9900 of the NRC Inspection Manual, 10 CFR Guidance for 10 CFR 50.59, Changes, Tests, and Experiments.


====b. Findings====
====b. Findings====
No findings of significance were identified.
No findings of significance were identified. {{a|1R17}}
{{a|1R17}}
==1R17 Permanent Plant Modifications==
==1R17 Permanent Plant Modifications==
{{IP sample|IP=IP 71111.17B}}
{{IP sample|IP=IP 71111.17B}}
.1Review of Permanent Plant Modifications
===.1 Review of Permanent Plant Modifications===


====a. Inspection Scope====
====a. Inspection Scope====
From March 6 through March 24, 2006, the inspectors reviewed twelve permanent plantmodifications that had been installed in the plant during the last two years. The modifications were chosen based upon risk significance, safety significance, and complexity. The inspectors reviewed the modifications to verify that the completed design changes were in accordance with the specified design requirements and the licensing bases and to confirm that the changes did not adversely affect any systems'safety function. Design and post-modification testing aspects were verified to ensure the functionality of the modification, its associated system, and any support systems. The inspectors also verified that the modifications performed did not place the plant in an increased risk configuration.
From March 6 through March 24, 2006, the inspectors reviewed twelve permanent plant modifications that had been installed in the plant during the last two years. The modifications were chosen based upon risk significance, safety significance, and complexity. The inspectors reviewed the modifications to verify that the completed design changes were in accordance with the specified design requirements and the licensing bases and to confirm that the changes did not adversely affect any systems' safety function. Design and post-modification testing aspects were verified to ensure the functionality of the modification, its associated system, and any support systems.
 
The inspectors also verified that the modifications performed did not place the plant in an increased risk configuration.
 
The inspectors also used applicable industry standards to evaluate acceptability of the modifications. The list of modifications and other documents reviewed by the inspectors is included as an attachment to this report.


The inspectors also used applicable industry standards to evaluate acceptability of themodifications. The list of modifications and other documents reviewed by the inspectors is included as an attachment to this report.The Prairie Island Unit 1 reactor vessel head replacement modification, which affectsthe barrier integrity cornerstone, was not selected as part of this inspection. This modification will be inspected at a later date in accordance with inspection procedure71007, "Reactor Vessel Head Replacement Inspection."
The Prairie Island Unit 1 reactor vessel head replacement modification, which affects the barrier integrity cornerstone, was not selected as part of this inspection. This modification will be inspected at a later date in accordance with inspection procedure 71007, Reactor Vessel Head Replacement Inspection.


====b. Findings====
====b. Findings====
b.1Failure to Consider Adverse Ampacity Effects of High Ambient Temperature Conditionsin the Auxiliary Feedwater Pump RoomsIntroduction: On March 15, 2006, the inspectors identified a Non-Cited Violation of10 CFR Part 50, Appendix B Criterion III, "Design Control," of very low safetysignificance (Green). Specifically, the licensee had not evaluated and updated the associated plant cable ampacity calculation to determine the potential consequences ofadverse effects to cabling due to higher temperatures in the AFW pump rooms and other auxiliary building areasDiscussion: Revised licensee calculation ENG-ME-021, "Auxiliary Feedwater PumpRoom Heat-up," indicated that the potential maximum ambient temperature in the AFWpump rooms could reach up to 127F. The potential high ambient temperature couldoccur during post accident mitigation (an extended loss of off-site power) and when the initial ambient temperature in the rooms was at 104F. The licensee evaluated the effects of the high ambient temperature on the safety-relatedequipment (i.e., motor driven AFW pump motors, motor operated valves, motor control centers, transformers and hot shutdown panel) located in the AFW pump rooms. The evaluation was documented in calculation ENG-ME-021, Revision 2 and concluded that the operability of the safety-related equipment located in the rooms was acceptable foran ambient room temperature of 127F. This conclusion was also documented in thelicensee's 10 CFR 50.59 screening number 2469, Revision 0. However, the licensee failed to address the effects of these heightened temperatures on the ampacity of electrical cables in the rooms. The inspectors also reviewed the licensee's 10 CFR 50.59 Safety Evaluation Number 1037 "Affect of Revised Unit 1 Main Steam Line Break on Auxiliary Building Environment,"  which identified that the ambient temperature couldalso reach up to 122F in several areas in the auxiliary building. Prairie Island Engineering Manual for Electrical Cables Design, Fabrication andInstallation Summary was based on an ambient temperature of 104F. Other plantspecific evaluations (i.e. Calculation ENG-EE-019 and Safety Evaluation 369) which have previously evaluated potential cable ampacity issues were also based on an ambient temperature of 104F. The licensee failed to evaluate and update the cableampacity calculation to evaluate the effects of potential high ambient temperatures onthe ampacity of electrical cables located in these rooms. Since higher temperatures adversely affect the ampacity of electrical cables, the higher temperatures in the AFW pump rooms and other plant areas had the potential to adversely affect the functionality and/or operability of equipment important to safety fed by cabling in these rooms. The inspectors were concerned that the possibility existed that some of the equipment fed bycables located in these areas may not function due to possible faulting of the supply cables. As a result of the inspectors' concerns, the licensee issued Action Request CAP 01018612. After performing a preliminary evaluation that assessed cabling in AFW pump rooms and the auxiliary building areas, the licensee determined that there was no evidence thatsafety related structures, systems, and components would not function as required. While the higher temperatures decreased the ampacity margins for the affected cabling, the licensee preliminarily determined that the margins it did not decrease to the limitwhere the cabling would fail if called upon to provide power to equipment important to safety.Analysis: The inspectors determined that this issue was a performance deficiencywarranting a significance evaluation, since the licensee failed to account for high temperature conditions in the AFW pump rooms and other several rooms located in the auxiliary building that adversely affected cables supplying power to equipment importantto safety. The finding was greater than minor in accordance with IMC 0612, "Power ReactorInspection Reports," Appendix B, "Issue Screening," because it affected the mitigating system cornerstone objective to ensure the availability, reliability, and capability ofsystems that mitigate transients and accidents, and if left uncorrected, the finding couldbecome a more significant safety concern. Specifically, if left uncorrected, the licensee may not account for high temperature conditions in plant areas that could adversely affect the ampacity of cabling that supply power to equipment important to safety. The inspectors determined the finding was of very low significance (Green) usingIMC 0609, Appendix A, "Significance Determination of Reactor Inspection Findings for the At-Power Situations," because the inspectors answered "no" to all five questions under the Mitigating Systems Cornerstone column of the Phase 1 worksheet. In particular, the licensee's preliminary evaluation determined that the higher temperaturesin the AFW pump rooms and other auxiliary building areas would not prevent equipmentimportant to safety from functioning.
b.1 Failure to Consider Adverse Ampacity Effects of High Ambient Temperature Conditions    in the Auxiliary Feedwater Pump Rooms
 
=====Introduction:=====
On March 15, 2006, the inspectors identified a Non-Cited Violation of 10 CFR Part 50, Appendix B Criterion III, Design Control, of very low safety significance (Green). Specifically, the licensee had not evaluated and updated the associated plant cable ampacity calculation to determine the potential consequences of adverse effects to cabling due to higher temperatures in the AFW pump rooms and other auxiliary building areas Discussion: Revised licensee calculation ENG-ME-021, Auxiliary Feedwater Pump Room Heat-up, indicated that the potential maximum ambient temperature in the AFW pump rooms could reach up to 127EF. The potential high ambient temperature could occur during post accident mitigation (an extended loss of off-site power) and when the initial ambient temperature in the rooms was at 104EF.
 
The licensee evaluated the effects of the high ambient temperature on the safety-related equipment (i.e., motor driven AFW pump motors, motor operated valves, motor control centers, transformers and hot shutdown panel) located in the AFW pump rooms. The evaluation was documented in calculation ENG-ME-021, Revision 2 and concluded that the operability of the safety-related equipment located in the rooms was acceptable for an ambient room temperature of 127EF. This conclusion was also documented in the licensees 10 CFR 50.59 screening number 2469, Revision 0. However, the licensee failed to address the effects of these heightened temperatures on the ampacity of electrical cables in the rooms. The inspectors also reviewed the licensees 10 CFR 50.59 Safety Evaluation Number 1037 Affect of Revised Unit 1 Main Steam Line Break on Auxiliary Building Environment, which identified that the ambient temperature could also reach up to 122EF in several areas in the auxiliary building.
 
Prairie Island Engineering Manual for Electrical Cables Design, Fabrication and Installation Summary was based on an ambient temperature of 104EF. Other plant specific evaluations (i.e. Calculation ENG-EE-019 and Safety Evaluation 369) which have previously evaluated potential cable ampacity issues were also based on an ambient temperature of 104EF. The licensee failed to evaluate and update the cable ampacity calculation to evaluate the effects of potential high ambient temperatures on the ampacity of electrical cables located in these rooms. Since higher temperatures adversely affect the ampacity of electrical cables, the higher temperatures in the AFW pump rooms and other plant areas had the potential to adversely affect the functionality and/or operability of equipment important to safety fed by cabling in these rooms. The
 
inspectors were concerned that the possibility existed that some of the equipment fed by cables located in these areas may not function due to possible faulting of the supply cables. As a result of the inspectors concerns, the licensee issued Action Request CAP 01018612.
 
After performing a preliminary evaluation that assessed cabling in AFW pump rooms and the auxiliary building areas, the licensee determined that there was no evidence that safety related structures, systems, and components would not function as required.
 
While the higher temperatures decreased the ampacity margins for the affected cabling, the licensee preliminarily determined that the margins it did not decrease to the limit where the cabling would fail if called upon to provide power to equipment important to safety.
 
=====Analysis:=====
The inspectors determined that this issue was a performance deficiency warranting a significance evaluation, since the licensee failed to account for high temperature conditions in the AFW pump rooms and other several rooms located in the auxiliary building that adversely affected cables supplying power to equipment important to safety.
 
The finding was greater than minor in accordance with IMC 0612, Power Reactor Inspection Reports, Appendix B, Issue Screening, because it affected the mitigating system cornerstone objective to ensure the availability, reliability, and capability of systems that mitigate transients and accidents, and if left uncorrected, the finding could become a more significant safety concern. Specifically, if left uncorrected, the licensee may not account for high temperature conditions in plant areas that could adversely affect the ampacity of cabling that supply power to equipment important to safety.
 
The inspectors determined the finding was of very low significance (Green) using IMC 0609, Appendix A, Significance Determination of Reactor Inspection Findings for the At-Power Situations, because the inspectors answered no to all five questions under the Mitigating Systems Cornerstone column of the Phase 1 worksheet. In particular, the licensees preliminary evaluation determined that the higher temperatures in the AFW pump rooms and other auxiliary building areas would not prevent equipment important to safety from functioning.


=====Enforcement:=====
=====Enforcement:=====
10 CFR Part 50, Appendix B, Criterion III, "Design Control" states, in part,that measures shall be established to assure that applicable design basis are correctly translated into specifications, drawings, procedures and instructions. Contrary to the above, the licensee did not have a design basis calculation for cable ampacity that supported the high temperatures that the AFW pump rooms and other plant areas couldexperience. The Prairie Island calculation and engineering manual that did address cable ampacity were significantly less conservative, since temperatures of 104F wereassumed where temperatures in these areas could exceed 122F.Because this issue was of very low safety significance, not willful, and because it wasentered in the licensee's corrective action program as CAP 01018612, this violation is being treated as an NCV, consistent with Section VI.A of the NRC Enforcement Policy.
10 CFR Part 50, Appendix B, Criterion III, Design Control states, in part, that measures shall be established to assure that applicable design basis are correctly translated into specifications, drawings, procedures and instructions. Contrary to the above, the licensee did not have a design basis calculation for cable ampacity that supported the high temperatures that the AFW pump rooms and other plant areas could experience. The Prairie Island calculation and engineering manual that did address cable ampacity were significantly less conservative, since temperatures of 104EF were assumed where temperatures in these areas could exceed 122EF.
 
Because this issue was of very low safety significance, not willful, and because it was entered in the licensees corrective action program as CAP 01018612, this violation is being treated as an NCV, consistent with Section VI.A of the NRC Enforcement Policy.


(NCV 05000282/2006006-01; 05000306/2006006-01)  
(NCV 05000282/2006006-01; 05000306/2006006-01)


==OTHER ACTIVITIES (OA)==
==OTHER ACTIVITIES (OA)==
4OA2Identification and Resolution of Problems.1Routine Review of Condition Reports
{{a|4OA2}}
==4OA2 Identification and Resolution of Problems==
 
===.1 Routine Review of Condition Reports===


====a. Inspection Scope====
====a. Inspection Scope====
From March 6 through March 24, 2006, the inspectors Action Process documents that identified or were related to 10 CFR 50.59 evaluationsand permanent plant modifications. The inspectors reviewed these documents to evaluate the effectiveness of corrective actions related to permanent plant modifications and evaluations for changes, tests, or experiments issues. In addition, corrective action documents written on issues identified during the inspection were reviewed to verify adequate problem identification and incorporation of the problems into the correctiveaction system. The specific corrective action documents that were sampled and reviewed by the team are listed in the attachment to this report.
From March 6 through March 24, 2006, the inspectors Action Process documents that identified or were related to 10 CFR 50.59 evaluations and permanent plant modifications. The inspectors reviewed these documents to evaluate the effectiveness of corrective actions related to permanent plant modifications and evaluations for changes, tests, or experiments issues. In addition, corrective action documents written on issues identified during the inspection were reviewed to verify adequate problem identification and incorporation of the problems into the corrective action system. The specific corrective action documents that were sampled and reviewed by the team are listed in the attachment to this report.


====b. Findings====
====b. Findings====
Line 99: Line 151:


==OTHER ACTIVITIES==
==OTHER ACTIVITIES==
4OA6Meetings.1Exit MeetingThe inspectors presented the inspection results to Mr. T. Palmisano and others of thelicensee's staff, on March 24, 2006. Licensee personnel acknowledged the inspection results presented. Licensee personnel were asked to identify any documents, materials, or information provided during the inspection that were considered proprietary other than those returned. No additional proprietary information was identified.ATTACHMENT:
{{a|4OA6}}
==4OA6 Meetings==
 
===.1 Exit Meeting===
 
The inspectors presented the inspection results to Mr. T. Palmisano and others of the licensees staff, on March 24, 2006. Licensee personnel acknowledged the inspection results presented. Licensee personnel were asked to identify any documents, materials, or information provided during the inspection that were considered proprietary other than those returned. No additional proprietary information was identified.
 
ATTACHMENT:  


=SUPPLEMENTAL INFORMATION=
=SUPPLEMENTAL INFORMATION=


==KEY POINTS OF CONTACT==
==KEY POINTS OF CONTACT==
Licensee
Licensee
: [[contact::T. Palmisano]], Site Vice President
: [[contact::T. Palmisano]], Site Vice President
Line 110: Line 170:
: [[contact::S. Thomas]], Design Engineering Supervisor
: [[contact::S. Thomas]], Design Engineering Supervisor
: [[contact::L. Gunderson]], Mechanical Design Engineer
: [[contact::L. Gunderson]], Mechanical Design Engineer
: [[contact::C. Sansome]], Mechanical Design EngineerNuclear Regulatory Commission
: [[contact::C. Sansome]], Mechanical Design Engineer
Nuclear Regulatory Commission
: [[contact::J. Adams]], Senior Resident Inspector
: [[contact::J. Adams]], Senior Resident Inspector
Attachment
 
==ITEMS OPENED, CLOSED, AND DISCUSSED==
==ITEMS OPENED, CLOSED, AND DISCUSSED==


===Opened===
===Opened===
None.Opened and
 
===Closed===
None.
: 05000282/2006006-01;
 
: [[Closes finding::05000306/FIN-2006006-01]] NCVFailure to Consider Adverse Ampacity Effects of HighTemperature Conditions in the Auxiliary feedwaterPump RoomsDiscussedNone.  
===Opened and Closed===
: Attachment
: 05000282/2006006-01; NCV  Failure to Consider Adverse Ampacity Effects of High
: 05000306/2006006-01       Temperature Conditions in the Auxiliary feedwater Pump Rooms
 
===Discussed===
 
None.
 
==LIST OF DOCUMENTS REVIEWED==
==LIST OF DOCUMENTS REVIEWED==
The following is a list of licensee documents reviewed during the inspection, includingdocuments prepared by others for the licensee.
 
: Inclusion on this list does not imply that NRCinspectors reviewed the documents in their entirety, but rather, that selected sections orportions of the documents were evaluated as part of the overall inspection effort.
: Inclusion of a document in this list does not imply NRC acceptance of the document, unless specifically statedin the inspection report.IR02Evaluation of Changes, Tests, or Experiments (71111.02)10
: CFR 50.59 ScreeningsNo. 1691;
: ENG-ME-538, Structural Evaluation of Bolts on 11 and 21 Fan Coil UnitMotors; Revision 0No. 2056; Modification 04CT02, Revise Cooling Tower Undervoltage Relaying;Revision 0No. 2272; Permanent Plant Modification 04RC03 - Pressurizer PORV Block ValveReplacement, Addendum 2 to Westinghouse Stress Analysis 0951s, Revision 0No. 2301;
: ENG-ME-576, AFW Pump Minimum Acceptance Criteria, Revision 0
: No. 2303; USAR Input Item #05001; Revision 0
: No. 2307; Calculation
: ENG-ME-443, Revision 3, PCRs
: 20050890,
: 20050891,20050892,
: 20050893; Revision 0 No. 2350; Calculation
: ENG-CS-278, Seismic Qualification of Components in ComponentCooling System Pressure Boundary; Revision 0No. 2365; Calculation
: ENG-ME-615, Tube Plugging Limits for 21 Containment Fan CoilUnit; Revision 0No. 2370;
: SP 1450, 31 Battery Refueling Outage Discharge Test; Revision 0
: No. 2371; T-Mod 05T186; Revision 0
: No. 2443; Calculation
: ENG-ME-621,
: CV-31998 and
: CV-31999 Air Receiver Capacity;Revision 0No. 2452; Containment Spray Pump Discharge Check Valve Closure AcceptanceCriteria; Revision 0No. 2469; Calculation
: ENG-ME-021, Auxiliary Feedwater Pump Room Heat-up;Revision 0No. 2486; Modification No. 05SA02, Structural Calculations S-11164-039-01 andS-11164-039-02; Revision 1
: AttachmentNo. 2509; Design Change 05CL03:
: Cooling Water Pump Bearing Water, Part 1: Enhance Well Water Normal Supply to Safeguards Cooling Water Pump Bearings;
: Revision 0No. 2513;
: ENG-ME-576, AFW Pump Minimum Acceptance Criteria;
: ENG-ME-454,Pressure Drop between Steam Generator and Safety Valve; TCNs and PCRs for
: SP 1102, 1103, 2102 and 2103; Revision 0No. 2523;
: ENG-ME-646 Revision 0 Addendum 1, Reinforcing of Component CoolingHeat Exchanger Divider Plate; Revision 010
: CFR 50.59 EvaluationsNo. 1025; Zebra Mussel Treatment; Revision 1 dated April 22, 2005
: No. 1032; Revised Containment Integrity Analysis with New Mass and Energy Methods;Revision 0; dated November 4, 2004No. 1035; Compensated Hi-Tavg Parameter Changes (TM-0401H); Revision 0; datedJanuary 20, 2005No. 1037; Affected Revised Unit 1 Main Steam Line Break on Auxiliary BuildingEnvironment; Revision 0; dated October 24, 2004No. 1038; Use of Ultimate Strength Design Methodology to Evaluate Vertical SeismicLoads on Floors; Revision 0 dated November 4, 2005No. 1046; Unit 2 Cycle 23 Core Reload; Revision 1 dated May 20, 2005
: No. 1047; Changes to Primary Chemistry Program Lithium and Hydrogen Limits;Revision 0 dated May 18, 2005No. 1050; Revised Small Break LOCA Analysis Using the NORTUMP Code, SI into theBroken Loop and COSI Condensation Model (WCAP-10054-P-A Add. 2 Revision. 1);
: Revision 0 dated January 13, 2006IR17Permanent Plant Modifications (71111.17B)ModificationsEEC No. 1378; Generic Change from Carbon Steel Globe Hancock Valves to CarbonSteel Globe Vogt Valves; Revision 0EEC No. 1503; Replace
: SV-33535; Revision 0
: EEC No. 1576; Upgrade CC HX TCV Positioners and F/Rs; Revision 0
: EEC No. 1616; Replace Breaker 121B-31 THEF MCCB with a THED; Revision 0
: EEC No. 1618; Replace 600 lb Valves with 800 lb Vogt Valves; Revision 0
: AttachmentEEC No. 1636; Longer Bolts for SI Accumulator Hangers and As Found Condition; Revision 004RC04; PRT Level Transmitter Replacement; Revision 0
: 05CL03; Enhance Well Water Normal Supply to Safeguards Cooling Water PumpBearings; Revision 005SA02; Replacement of No. 121 and No. 122 Instrument Air Dryers; Revision 0
: 05ST01; CT Underground Cable Replacement; dated August 31, 2005
: 1TM-401H; Change Operating Parameters on 1TM-401H; dated February 04, 2005
: 2TM-401H; Change Operating Parameters on 2TM-401H; dated February 07, 2005 
===Other Documents===
: Reviewed During InspectionCorrective Action Program Documents Generated As a Result of InspectionCAP
: 01018063; Acceptance Criteria in
: SP 1450 and 2450 Need Review; datedMarch 9, 2006CAP
: 01018337; Clarity Regarding Application of 50.59 to IST Acceptance Criteria;dated March 13, 2006CAP
: 01018612; Cable Ampacities Have Not Considered Increased AmbientTemperatures; dated March 15, 2006CAP
: 01019410;
: EEC 1378 Is Not Clear Regarding Not Using for Throttle ValveReplacement; dated March 20, 2006CAP
: 01019730; USAR Section 8.5.2 Needs to Be Clarified; dated March 22, 2006
: CAP 01019811;
: EEC 1636 Does Not Address Potential Reduction in Hanger Capacity;dated March 22, 2006CAP
: 01019822; Control of U-Bolt Configurations During Modifications; datedMarch 22, 2006CAP
: 01019883; DC Battery
: SP 2314 Contains Incorrect Data; dated March 22, 2006
: CAP 01020014; 50.59 Evaluation 1037 Does Not Clearly Address Input Parameters;dated March 23, 2006CAP
: 01020123; Errors on Logic Diagram; dated March 23, 2006
: AttachmentCorrective Action Program Documents Reviewed During the Inspection
: CAP 0039552; Auxiliary Building HELB Analysis Temperature Assumptions, datedOctober 28, 2004
: CAP 0078984; RC System Head Vent System - Design Configuration; datedDecember 20, 2004CAP
: 00841009; 50.59 Evaluation Bypassed Two Reviews Required by the ControllingAWI; dated March 10, 2005CAP
: 00843566; Change in Primary Lithium/PH Control; dated May 10, 2005
: CAP 00846108; Unable to Complete Repairs on CL Valves Due to Valve ConfigurationIssues; dated May 17, 2005CAP
: 00851046; The Design Basis for the Air Receiver for
: CV-31998 and
: CV-31999 isUnclear; dated May 28, 2005CAP
: 00854365; New Security Fencing Installed Without Apparent Design ChangeControls; dated June 7, 2005CAP
: 00884715; Temporary Cooling Added to
: TP 1636 and 1637 Needs a 50.59Screening; dated September 8, 2005CAP
: 00888817; 50.59 Process Self-Assessment for Screening No. 2042; datedSeptember 21, 2005CAP
: 00889055; 50.59 Process Self-Assessment Finding for Screening No. 2350; datedSeptember 21, 2005
: CAP 01017232; Configuration Management Self Assessment Finding - Modification05CL03; March 3, 2006CAP
: 01017244;
: EEC 1559- Replace the Pressurizer Porv Accumulator Air Inlet CheckValve; dated March 03, 2006CAP
: 01017281; Setpoint Calculations have not been Screened by 50.59 Process; datedMarch 03, 2006CalculationsENG-CS-278; Seismic Qualification of Components in Component Cooling SystemPressure Boundary; Revision 1ENG-EE-019; Evaluation to Resolve Overfilled Cable Trays Identified in Follow on ItemA0457; Revision 0ENG-ME-021; Auxiliary Feedwater Pump Room Heat-Up; Revision 2
: AttachmentENG-ME-454; Pressure Drop between SG and Safety Valve; Revision 0, Addendum 1ENG-ME-538; Structural Evaluation of Bolts on 11 and 21 FCU Motors; Revision 0
: ENG-ME-576; AFW Pump Minimum Acceptance Criteria; Revision 1
: ENG-ME-577; Prairie Island Characterization of Zebra Mussel Transport in Pump IntakeStructures; Revision 0ENG-ME-605; Prairie Island Zebra Mussel Transport Shell Deposition as a Function ofMussel Concentration; Revision 0ENG-ME-615; Tube Plugging Limits for 21 Containment Fan Coil Unit; Revision 0
: ENG-ME-621;
: CV-31998 and
: CV-31999 Air Receiver Capacity; Revision 0
: ENG-ME-646; Reinforcing of CC Hx Divider Plate - Vendor Calculation
: PI-S-021;Revision 0, Addendum 1PI-605044-P01; Evaluation of Well / Filtered Water Piping Below Elevation 695Screenhouse; Revision 0S-11164-039-01; Design of foundations for Instrument Air Dryers Nos. 121 and 122;Revision 0S-11164-039-02; Design of New Supports - Modification No. 05SA02; Revision 0
: S-B01-VS-001; Structural Floor Analysis for Vertical Seismic; Revision 0
: DrawingsFC-64-247A; Instrument Connection at
: LT 24060, Accumulator No. 22; Revision 4
: FC-64-250; Instrument Piping Connection at
: LT 24058, Accumulator No. 21; Revision 4
: ND-211543; 12,121 and 122 Cooling Water Pumps, Bearing Water Supply Isometric;Revision 0ND-211544; 12,121 and 122 Cooling Water Pumps, Bearing Water Supply PipingSupports; Revision 0NE-40009; Sheer 97.2; 11 TD Aux. Feedwater Pump Main Steam Supply ValveCV-31998; Revision
: DTNF-40312-1; Interlock Logic Diagram, Aux. Feedwater System - Unit 1; Revision
: ACX-HIAW-1106-6320; Pipe Support - Safety Injection; Revision
: AX-HIAW-1106-6324; Pipe Support - Safety Injection; Revision A
: AttachmentProceduresFP-E-EQV-01; Fleet Procedure; Equivalency Evaluations and Changes; Revision 0FP-SC-GEN-02; Fleet Procedure; Requesting Materials; Revision 6
: FP-SC-GEN-03; Fleet Procedure; Catalog Item Creation and Change; Revision 4
: FP-WM-PLA-01; Fleet Procedure:
: Work Order Planning Process; Revision 0
: SP 1353A:
: Surveillance Procedure; Quarterly Testing of
: CS-16 and
: CS-18, 11 CSPSuction and Discharge Check Valves; Revision 10SP 2103:
: Surveillance Procedure; 22 Turbine-Driven Auxiliary Feedwater Pump OnceEvery Refueling Shutdown Flow Test; Revision 42Miscellaneous DocumentsAction Request No.
: 01015987; USAR Change for Modification 05SA02 - Instrument AirDryer; dated February 23, 2006PCR 2005-1849B; Procedure Change Request,
: SP 1353A Revision 8; Quarterly Testingof
: CS-16 and
: CS-18, 11 CSP Suction and Discharge Check Valves; dated August 12, 2005PCR 2005-2881A; Procedure Change Request,
: SP 2103 Revision 41;22 Turbine-DrivenAuxiliary Feedwater Pump Once Every Refueling Shutdown Flow Test; datedOctober 27, 2005Safety Evaluation No. 342; Place CT BT112 in Manual for all Operating Condition; datedMarch 12, 1993Safety Evaluation No. 369; Cable Tray F
ill and Spacing Concerns; Revision 0; datedAugust 25, 1995Safety Evaluation No. 478-A1-04; USAR Up-date Appendix I.11 (Compartment Pressureand Temperatures); Revision 0; dated February 24, 2000
: SP-2314; 22 Battery Refueling Outage Discharge Test; Performed on May 19, 2005
: TCN 2005-1104; Temporary Change Notice,
: SP 1353A Revision 8: Quarterly Testing ofCS-16 and
: CS-18, 11 CSP Suction and Discharge Check Valves; dated July 18, 2005TCN 2005-1117; Temporary Change Notice,
: SP 1353A Revision 8: Quarterly Testing ofCS-16 and
: CS-18, 11 CSP Suction and Discharge Check Valves; dated July 22, 2005
: Attachment
==LIST OF ACRONYMS==
USEDADAMSAgency-Wide Document Access and Management SystemAFWAuxiliary FeedwaterCFRCode of Federal Regulations
DRSDivision of Reactor Safety
EECEquivalent Engineering Change
NCVNon-Cited Violation
NEINuclear Energy Institute
NRCNuclear Regulatory Commission
PCRProcedure Change Request
SDPSignificance Determination Process
TCNTemporary Change Notice
: [[USARU]] [[pdated Safety Analysis Report]]
}}
}}

Latest revision as of 14:30, 22 December 2019

IR 05000282-06-006 (Drs); 05000306-06-006 (Drs); 03/06/2006 - 03/24/2006; Prairie Island Nuclear Generating Plant, Units 1 and 2; Evaluation of Changes, Tests, or Experiments (10 CFR 50.59) and Permanent Plant Modifications
ML061220751
Person / Time
Site: Prairie Island  Xcel Energy icon.png
Issue date: 04/28/2006
From: Dave Hills
NRC/RGN-III/DRS/EB1
To: Thomas J. Palmisano
Nuclear Management Co
References
IR-06-006
Download: ML061220751 (19)
v • d • e


Text

SUBJECT:

PRAIRIE ISLAND NUCLEAR GENERATING PLANT, UNITS 1 AND 2 NRC EVALUATION OF CHANGES, TESTS, OR EXPERIMENTS AND PERMANENT PLANT MODIFICATIONS BASELINE INSPECTION REPORT 05000282/2006006 (DRS); 05000306/2006006 (DRS)

Dear Mr. Palmisano:

On March 24, 2006, the U.S. Nuclear Regulatory Commission (NRC) completed a combined baseline inspection of the Evaluation of Changes, Tests, or Experiments and Permanent Plant Modifications at the Prairie Island Nuclear Generating Plant. The enclosed report documents the results of the inspection, which were discussed and others of your staff at the completion of the inspection on March 24, 2006.

The inspectors examined activities conducted under your license as they relate to safety and compliance with the Commissions rules and regulations and with the conditions of your license.

The inspectors reviewed selected procedures and records, observed activities, and interviewed personnel.

Based on the results of the inspection, one NRC-identified finding of very low safety significance was identified which involved a violation of NRC requirements. However, because this violation was of very low safety significance, not willful, and because it was entered into your corrective action program, the NRC is treating the issue as a Non-Cited Violation in accordance with Section VI.A.1 of the NRCs Enforcement Policy.

If you contest the subject or severity of a Non-Cited Violation, 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 Prairie Island Nuclear Generating Plant facility.

In accordance with 10 CFR 2.390 of the NRC's "Rules of Practice," a copy of this letter and its enclosure will be available electronically for public inspection in the NRC Public Document Room or from the Publicly Available Records (PARS) component of NRC's document system (ADAMS). ADAMS is accessible from the NRC Web site at http://www.nrc.gov/reading-rm/adams.html (the Public Electronic Reading Room).

Sincerely,

/RA/

David E. Hills, Chief Engineering Branch 1 Division of Reactor Safety Docket Nos. 50-282; 50-306 License Nos.

Enclosure:

Inspection Report 05000282/2006006 (DRS); 05000306/2006006 (DRS)

Document Room or from the Publicly Available Records (PARS) component of NRC's document system (ADAMS). ADAMS is a

REGION III==

Docket No: 50-282; 50-306 License No:

Report No: 05000282/2006006 (DRS); 05000306/2006006 (DRS)

Licensee:

Facility: Prairie Island Nuclear Generating Plant Location:

Dates: March 6 through March 24, 2006 Inspectors: J. Neurauter, Senior Reactor Inspector, Team Leader Alan Dahbur, Reactor Inspector Approved by: D. Hills, Chief Engineering Branch 1 Division of Reactor Safety (DRS)

SUMMARY OF FINDINGS

IR 05000282/2006006 (DRS); 05000306/2006006 (DRS); 03/06/2006 - 03/24/2006; Prairie

Island Nuclear Generating Plant, Units 1 and 2; Evaluation of Changes, Tests, or Experiments (10 CFR 50.59) and Permanent Plant Modifications.

The inspection covered a two-week announced baseline inspection on evaluations of changes, tests, or experiments and permanent plant modifications. The inspection was conducted by two regional based engineering inspectors. One Green Non-Cited Violation (NCV) was identified.

The significance of most findings is indicated by their color (Green, White, Yellow, Red), using Inspection Manual Chapter 0609, Significance Determination Process (SDP.) Findings for which the SDP does not apply, may be Green, or be assigned a severity level after NRC management review. The NRC's program for overseeing the safe operation of commercial nuclear power reactors is described in NUREG-1649, Reactor Oversight Process, Revision 3, dated July 2000.

A. Inspector-Identified and Self-Revealed Findings

Cornerstone: Mitigating Systems

Green.

A Non-Cited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, having very low safety significance was identified by the inspectors.

Specifically, the licensee had not evaluated and updated the associated plant cable ampacity calculation to determine the potential consequences of adverse effects to cabling due to higher temperatures in the auxiliary feedwater (AFW) pump rooms and other auxiliary building areas. After identification by the inspectors, the licensee was able to demonstrate that even though the higher temperatures decreased the ampacity margins for the affected cabling, it did not decrease the margins to the limit where the cabling would fail if called upon to provide power to equipment important to safety.

The finding was more than minor because it affected the mitigating system cornerstone objective to ensure the availability, reliability, and capability of systems that mitigate transients and accidents, and if left uncorrected, the finding could become a more significant safety concern. Specifically, if left uncorrected, the licensee may not account for high temperature conditions in plant areas that could adversely affect the ampacity of cabling that supply power to equipment important to safety. This finding was of very low safety significance because, the licensees preliminary evaluation determined that the higher temperatures in the AFW pump rooms and other auxiliary building areas would not prevent equipment important to safety from functioning. (Section 1R17.1.b.1)

Cornerstone: Barrier Integrity

No findings of significance were identified.

Licensee-Identified Violations

None.

REPORT DETAILS

REACTOR SAFETY

Cornerstones: Initiating Events, Mitigating Systems, and Barrier Integrity

1R02 Evaluations of Changes, Tests, or Experiments

Main article: IP 71111.02

.1 Review of 10 CFR 50.59 Evaluations and Screenings

a. Inspection Scope

From March 6 through March 24, 2006, the inspectors reviewed eight evaluations performed pursuant to 10 CFR 50.59. The inspectors confirmed that the evaluations were thorough and that prior NRC approval was obtained as appropriate. The inspectors also reviewed seventeen screenings where licensee personnel had determined that a 10 CFR 50.59 evaluation was not necessary. In regard to the changes reviewed where no 10 CFR 50.59 evaluation was performed, the inspectors verified that the changes did not meet the threshold to require a 10 CFR 50.59 evaluation. The evaluations and screenings were chosen based on risk significance, safety significance, and complexity. The list of documents reviewed by the inspectors is included as an attachment to this report.

The inspectors used, in part, Nuclear Energy Institute (NEI) 96-07, Guidelines for 10 CFR 50.59 Implementation, Revision 1, to determine acceptability of the completed evaluations and screenings. The NEI document was endorsed by the NRC in Regulatory Guide 1.187, Guidance for Implementation of 10 CFR 50.59, Changes, Tests, and Experiments, dated November 2000. The inspectors also consulted Part 9900 of the NRC Inspection Manual, 10 CFR Guidance for 10 CFR 50.59, Changes, Tests, and Experiments.

b. Findings

No findings of significance were identified.

1R17 Permanent Plant Modifications

Main article: IP 71111.17B

.1 Review of Permanent Plant Modifications

a. Inspection Scope

From March 6 through March 24, 2006, the inspectors reviewed twelve permanent plant modifications that had been installed in the plant during the last two years. The modifications were chosen based upon risk significance, safety significance, and complexity. The inspectors reviewed the modifications to verify that the completed design changes were in accordance with the specified design requirements and the licensing bases and to confirm that the changes did not adversely affect any systems' safety function. Design and post-modification testing aspects were verified to ensure the functionality of the modification, its associated system, and any support systems.

The inspectors also verified that the modifications performed did not place the plant in an increased risk configuration.

The inspectors also used applicable industry standards to evaluate acceptability of the modifications. The list of modifications and other documents reviewed by the inspectors is included as an attachment to this report.

The Prairie Island Unit 1 reactor vessel head replacement modification, which affects the barrier integrity cornerstone, was not selected as part of this inspection. This modification will be inspected at a later date in accordance with inspection procedure 71007, Reactor Vessel Head Replacement Inspection.

b. Findings

b.1 Failure to Consider Adverse Ampacity Effects of High Ambient Temperature Conditions in the Auxiliary Feedwater Pump Rooms

Introduction:

On March 15, 2006, the inspectors identified a Non-Cited Violation of 10 CFR Part 50, Appendix B Criterion III, Design Control, of very low safety significance (Green). Specifically, the licensee had not evaluated and updated the associated plant cable ampacity calculation to determine the potential consequences of adverse effects to cabling due to higher temperatures in the AFW pump rooms and other auxiliary building areas Discussion: Revised licensee calculation ENG-ME-021, Auxiliary Feedwater Pump Room Heat-up, indicated that the potential maximum ambient temperature in the AFW pump rooms could reach up to 127EF. The potential high ambient temperature could occur during post accident mitigation (an extended loss of off-site power) and when the initial ambient temperature in the rooms was at 104EF.

The licensee evaluated the effects of the high ambient temperature on the safety-related equipment (i.e., motor driven AFW pump motors, motor operated valves, motor control centers, transformers and hot shutdown panel) located in the AFW pump rooms. The evaluation was documented in calculation ENG-ME-021, Revision 2 and concluded that the operability of the safety-related equipment located in the rooms was acceptable for an ambient room temperature of 127EF. This conclusion was also documented in the licensees 10 CFR 50.59 screening number 2469, Revision 0. However, the licensee failed to address the effects of these heightened temperatures on the ampacity of electrical cables in the rooms. The inspectors also reviewed the licensees 10 CFR 50.59 Safety Evaluation Number 1037 Affect of Revised Unit 1 Main Steam Line Break on Auxiliary Building Environment, which identified that the ambient temperature could also reach up to 122EF in several areas in the auxiliary building.

Prairie Island Engineering Manual for Electrical Cables Design, Fabrication and Installation Summary was based on an ambient temperature of 104EF. Other plant specific evaluations (i.e. Calculation ENG-EE-019 and Safety Evaluation 369) which have previously evaluated potential cable ampacity issues were also based on an ambient temperature of 104EF. The licensee failed to evaluate and update the cable ampacity calculation to evaluate the effects of potential high ambient temperatures on the ampacity of electrical cables located in these rooms. Since higher temperatures adversely affect the ampacity of electrical cables, the higher temperatures in the AFW pump rooms and other plant areas had the potential to adversely affect the functionality and/or operability of equipment important to safety fed by cabling in these rooms. The

inspectors were concerned that the possibility existed that some of the equipment fed by cables located in these areas may not function due to possible faulting of the supply cables. As a result of the inspectors concerns, the licensee issued Action Request CAP 01018612.

After performing a preliminary evaluation that assessed cabling in AFW pump rooms and the auxiliary building areas, the licensee determined that there was no evidence that safety related structures, systems, and components would not function as required.

While the higher temperatures decreased the ampacity margins for the affected cabling, the licensee preliminarily determined that the margins it did not decrease to the limit where the cabling would fail if called upon to provide power to equipment important to safety.

Analysis:

The inspectors determined that this issue was a performance deficiency warranting a significance evaluation, since the licensee failed to account for high temperature conditions in the AFW pump rooms and other several rooms located in the auxiliary building that adversely affected cables supplying power to equipment important to safety.

The finding was greater than minor in accordance with IMC 0612, Power Reactor Inspection Reports, Appendix B, Issue Screening, because it affected the mitigating system cornerstone objective to ensure the availability, reliability, and capability of systems that mitigate transients and accidents, and if left uncorrected, the finding could become a more significant safety concern. Specifically, if left uncorrected, the licensee may not account for high temperature conditions in plant areas that could adversely affect the ampacity of cabling that supply power to equipment important to safety.

The inspectors determined the finding was of very low significance (Green) using IMC 0609, Appendix A, Significance Determination of Reactor Inspection Findings for the At-Power Situations, because the inspectors answered no to all five questions under the Mitigating Systems Cornerstone column of the Phase 1 worksheet. In particular, the licensees preliminary evaluation determined that the higher temperatures in the AFW pump rooms and other auxiliary building areas would not prevent equipment important to safety from functioning.

Enforcement:

10 CFR Part 50, Appendix B, Criterion III, Design Control states, in part, that measures shall be established to assure that applicable design basis are correctly translated into specifications, drawings, procedures and instructions. Contrary to the above, the licensee did not have a design basis calculation for cable ampacity that supported the high temperatures that the AFW pump rooms and other plant areas could experience. The Prairie Island calculation and engineering manual that did address cable ampacity were significantly less conservative, since temperatures of 104EF were assumed where temperatures in these areas could exceed 122EF.

Because this issue was of very low safety significance, not willful, and because it was entered in the licensees corrective action program as CAP 01018612, this violation is being treated as an NCV, consistent with Section VI.A of the NRC Enforcement Policy.

(NCV 05000282/2006006-01; 05000306/2006006-01)

OTHER ACTIVITIES (OA)

4OA2 Identification and Resolution of Problems

.1 Routine Review of Condition Reports

a. Inspection Scope

From March 6 through March 24, 2006, the inspectors Action Process documents that identified or were related to 10 CFR 50.59 evaluations and permanent plant modifications. The inspectors reviewed these documents to evaluate the effectiveness of corrective actions related to permanent plant modifications and evaluations for changes, tests, or experiments issues. In addition, corrective action documents written on issues identified during the inspection were reviewed to verify adequate problem identification and incorporation of the problems into the corrective action system. The specific corrective action documents that were sampled and reviewed by the team are listed in the attachment to this report.

b. Findings

No findings of significance were identified.

OTHER ACTIVITIES

4OA6 Meetings

.1 Exit Meeting

The inspectors presented the inspection results to Mr. T. Palmisano and others of the licensees staff, on March 24, 2006. Licensee personnel acknowledged the inspection results presented. Licensee personnel were asked to identify any documents, materials, or information provided during the inspection that were considered proprietary other than those returned. No additional proprietary information was identified.

ATTACHMENT:

SUPPLEMENTAL INFORMATION

KEY POINTS OF CONTACT

Licensee

T. Palmisano, Site Vice President
C. Mundt, Design Engineering Manager
J. Kivi, Senior Regulatory Compliance Engineer
S. Thomas, Design Engineering Supervisor
L. Gunderson, Mechanical Design Engineer
C. Sansome, Mechanical Design Engineer

Nuclear Regulatory Commission

J. Adams, Senior Resident Inspector

ITEMS OPENED, CLOSED, AND DISCUSSED

Opened

None.

Opened and Closed

05000282/2006006-01; NCV Failure to Consider Adverse Ampacity Effects of High
05000306/2006006-01 Temperature Conditions in the Auxiliary feedwater Pump Rooms

Discussed

None.

LIST OF DOCUMENTS REVIEWED

Retrieved from "https://kanterella.com/w/index.php?title=IR_05000282/2006006&oldid=2286546"