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| issue date = 02/17/2015
| issue date = 02/17/2015
| title = IR 05000346/2014007; on 11/03/2014 - 01/09/2015; Davis-Besse Nuclear Power Station; Evaluations of Changes, Tests, and Experiments and Permanent Plant Modifications
| title = IR 05000346/2014007; on 11/03/2014 - 01/09/2015; Davis-Besse Nuclear Power Station; Evaluations of Changes, Tests, and 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 = Lieb R
| addressee name = Lieb R
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=Text=
=Text=
{{#Wiki_filter:UNITED STATES NUCLEAR REGULATORY COMMISSION REGION III 2443 WARRENVILLE RD. SUIT E 210 LISLE, IL 60532
{{#Wiki_filter:UNITED STATES ary 17, 2015
-4352 February 1 7, 2015 Mr. Raymond Lieb Site Vice President FirstEnergy Nuclear Operating Co.


Davis-Besse Nuclear Power Station 5501 N. State Rte 2, Mail Stop A
==SUBJECT:==
-DB-3080 Oak Harbor, OH 43449
DAVIS-BESSE NUCLEAR POWER STATION EVALUATIONS OF CHANGES, TESTS, AND EXPERIMENTS AND PERMANENT PLANT MODIFICATIONS BASELINE INSPECTION REPORT 05000346/2014007
-9760
 
SUBJECT: DAVIS-BESSE NUCLEAR POWER STATION EVALUATION S OF CHANGES, TESTS, AND EXPERIMENTS AND PERMANENT PLANT MODIFICATIONS BASELINE INSPECTION REPORT 05000346/2014007


==Dear Mr. Lieb:==
==Dear Mr. Lieb:==
On January 9, 201 5 , the U.S. Nuclear Regulatory Commission (NRC) completed an Evaluations of Changes, Tests, and Experiments and Permanent Plant Modifications Inspection at your Davis-Besse Nuclear Power Station. The enclosed inspection report documents the inspection results which were discussed on January 9, 201 5, with you and other members of your staff.
On January 9, 2015, the U.S. Nuclear Regulatory Commission (NRC) completed an Evaluations of Changes, Tests, and Experiments and Permanent Plant Modifications Inspection at your Davis-Besse Nuclear Power Station. The enclosed inspection report documents the inspection results which were discussed on January 9, 2015, with you 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. One NRC-identified finding of very-low safety significance (Green) was identified during this inspection.
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.


This finding was determined to involve a violation of NRC requirements.
The inspectors reviewed selected procedures and records, observed activities, and interviewed personnel.


However, because of the very-low safety significance and because the iss ue was entered into your corrective action program, the NRC is treating the issue as a Non-Cited Violation (NCV) i n accordance with Section 2.3.2 of the NRC Enforcement Policy.
One NRC-identified finding of very-low safety significance (Green) was identified during this inspection. This finding was determined to involve a violation of NRC requirements. However, because of the very-low safety significance and because the issue was entered into your corrective action program, the NRC is treating the issue as a Non-Cited Violation (NCV) in accordance with Section 2.3.2 of the NRC Enforcement Policy.


If you contes t the violation or significance of the NCV
If you contest the violation or significance of the 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 copies to the Regional Administrator, U.S. Nuclear Regulatory Commission - Region III; the Director, Office of Enforcement, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001; and the Resident Inspector at Davis-Besse Nuclear Power 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 Davis-Besse Nuclear Power Station. In accordance with Title 10 of the Code of Federal Regulations (10 CFR) 2.390, Public Inspections, Exemptions, Requests for Withholding, 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 NRCs Public Document Room or from the Publicly Available Records (PARS)
, 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
component of the NRC's Agencywide Documents Access and Management System (ADAMS).
-0001; with copies to the Regional Administrator, U.S. Nuclear Regulatory Commission  
- Region III; the Director, Office of Enforcement, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001; and the Resident Inspector at Davis-Besse Nuclear Power 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 Davis-Besse Nuclear Power Station
. In accordance with Title 10 of the Code of Federal Regulations (10 CFR) 2.390, "Public Inspections, Exemptions, Requests for Withholding," 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's Public Document Room or from the Publicly Available Records (PARS) component of the NRC's Agencywide Documents Access and Management System (ADAMS).


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


Sincerely,
Sincerely,
/RA/
/RA/
Robert C. Daley, Chief Engineering Branch 3 Division of Reactor Safety Docket Nos.
Robert C. Daley, Chief Engineering Branch 3 Division of Reactor Safety Docket Nos. 50-346; 72-017 License No. NPF-3


50-346; 72-017 License No. NPF-3 Enclosure:
===Enclosure:===
Inspection Report 0500 0346/20 1 4007 w/Attachment: Supplemental Information cc w/encl:
Inspection Report 05000346/2014007 w/Attachment: Supplemental Information
Distribution via LISTSERV Enclosure U. S. NUCLEAR REGULATORY COMMISSION


==REGION III==
REGION III==
Docket No s: 50-346; 72-017 License No
Docket Nos: 50-346; 72-017 License No: NPF-3 Report No: 05000346/2014007 Licensee: FirstEnergy Nuclear Operating Company (FENOC)
: NPF-3 Report No:
Facility: Davis-Besse Nuclear Power Station Location: Oak Harbor, OH Dates: November 3, 2014 - January 09, 2015 Inspectors: Jasmine Gilliam, Acting Senior Reactor Inspector (Lead)
05000346/20 14 00 7 Licensee: FirstEnergy Nuclear Operating Company (FENOC) Facility: Davis-Besse Nuclear Power Station Location: Oak Harbor, OH Dates: November 3, 2014 - January 09, 2015 Inspectors:
Loyd Cain, Senior Resident Inspector (Vogtle)
Jasmine Gilliam, Acting Senior Reactor Inspector (Lead)
Ijaz Hafeez, Reactor Inspector Lionel Rodriguez, Reactor Inspector (Observer)
Loyd Cain, Senior Resident Inspector (Vogtle) Ijaz Hafeez, Reactor Inspector Lionel Rodriguez, Reactor Inspector (Observer)
Approved by: Robert C. Daley, Chief Engineering Branch 3 Division of Reactor Safety Enclosure
Approved by:
Robert C. Daley, Chief Engineering Branch 3 Division of Reactor Safety 2


=SUMMARY OF FINDINGS=
=SUMMARY OF FINDINGS=
IR 05000346/20 1400 7; 11/03/2014 - 01/09/2015; Davis-Besse Nuclear Power Station
IR 05000346/2014007; 11/03/2014 - 01/09/2015; Davis-Besse Nuclear Power Station;
; Evaluations of Changes, Tests, and Experiments and Permanent Plant Modifications.


This report covers a two-week announced baseline inspection on evaluations of changes, tests, and experiments and permanent plant modifications. The inspection was conducted by Region III based engineering inspectors.
Evaluations of Changes, Tests, and Experiments and Permanent Plant Modifications.


One finding of very-low safety significance was identified by the inspectors. The findin g was considered a Non-Cited Violation (NCV) of U.S. Nuclear Regulatory Commission (NRC)regulations. The significance of most findings is indicated by their color (i.e., greater than Green, or Green, White, Yellow, Red) using Inspection Manual Chapter (IMC) 0609, "Significance Determination Process" (SDP).
This report covers a two-week announced baseline inspection on evaluations of changes, tests, and experiments and permanent plant modifications. The inspection was conducted by Region III based engineering inspectors. One finding of very-low safety significance was identified by the inspectors. The finding was considered a Non-Cited Violation (NCV) of U.S. Nuclear Regulatory Commission (NRC) regulations. The significance of most findings is indicated by their color (i.e., greater than Green, or Green, White, Yellow, Red) using Inspection Manual Chapter (IMC) 0609, Significance Determination Process (SDP). Cross-cutting aspects were determined using IMC 0310, Aspects 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. All violations of NRC requirements are dispositioned in accordance with the NRCs Enforcement Policy dated July 9, 2013. The NRCs program for overseeing the safe operation of commercial nuclear power reactors is described in NUREG-1649, Reactor Oversight Process, Revision 5, dated February 2014.


Cross-cutting aspects were determined using IMC 0310, "Aspect s 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. All violations of NRC requirements are dispositioned in accordance with the NRC's Enforcement Policy dated J ul y 9, 201 3. The NRC's program for overseeing the safe operation of commercial nuclear power reactors is described in NUREG
===NRC-Identified===
-1649, "Reacto r Oversight Process," Revision 5, dated February 2014
and Self-Revealed Findings
. A. N RC-Identified and Self-Revealed Findings  


===Cornerstone: Initiating Events===
===Cornerstone: Initiating Events===
: '''Green.'''
: '''Green.'''
The inspectors identified a finding of very-low safety significance (Green) and associated Non-Cited Violation (NCV) of 10 CFR Part 50, Appendix B, Criterion III, "Design Control" for the licensee's failure to install and control the Reactor Coolant Pump (RCP) seal cavity vent flexible hose s per the design basis analysis. Specifically, the licensee failed to correctly translate the design basis installation configuration and installation fatigue analysis in calculation SP
The inspectors identified a finding of very-low safety significance (Green) and associated Non-Cited Violation (NCV) of 10 CFR Part 50, Appendix B, Criterion III,
-274-I, "Pipe Stress Analysis:
Design Control for the licensees failure to install and control the Reactor Coolant Pump (RCP) seal cavity vent flexible hoses per the design basis analysis. Specifically, the licensee failed to correctly translate the design basis installation configuration and installation fatigue analysis in calculation SP-274-I, Pipe Stress Analysis: Reactor Coolant Pump 1-1-1 Seal Cavity Vent, into specifications, drawings, procedures, and instructions. The licensee entered this finding into their Corrective Action Program (CAP) to review the lack of controls over the installation of the flexible hoses, but determined that the flexible hoses remained operable.
Reactor Coolant Pump 1
-1-1 Seal Cavity Vent
," into specifications, drawings, procedures, and instructions. The licensee entered this finding into their Corrective Action Program (CAP) to review the lack of control s over the installation of the flexible hoses, but determined that the flexible hoses remained operable.


The performance deficiency was determined to be more than minor because
The performance deficiency was determined to be more than minor because, if left uncorrected, it would have the potential to lead to a more significant safety concern.
, if left uncorrected
, it would have the potential to lead to a more significant safety concern. Specifically, the failure to install and control the flexible hose s in accordance with the design basis analysis could lead to failure of the hoses due to operation beyond their analyzed limits. The finding screened as of very-low safety significance (Green) because the finding could not result in exceeding the Reactor Coolant System (R CS) leak rate for a small Loss of Coolant Accident (LOCA) after a reasonable assessment of degradation, and it could not have likely affected other systems used to mitigate a LOCA resulting in a total loss of their function after a reasonable assessment of degradation. The inspectors determined this finding had an associated cross-cutting aspect, Design Margins, in the Human Performance cross-cutting area. Specifically, the licensee did not carefully guard and change the RCP seal cavity vent lines, which form part of the


RCS fission product barrier, through a systematic and rigorous process.
Specifically, the failure to install and control the flexible hoses in accordance with the design basis analysis could lead to failure of the hoses due to operation beyond their analyzed limits. The finding screened as of very-low safety significance (Green)because the finding could not result in exceeding the Reactor Coolant System (RCS)leak rate for a small Loss of Coolant Accident (LOCA) after a reasonable assessment of degradation, and it could not have likely affected other systems used to mitigate a LOCA resulting in a total loss of their function after a reasonable assessment of degradation.


[H.6] (Section 1R17.2.b.(1))  
The inspectors determined this finding had an associated cross-cutting aspect, Design Margins, in the Human Performance cross-cutting area. Specifically, the licensee did not carefully guard and change the RCP seal cavity vent lines, which form part of the RCS fission product barrier, through a systematic and rigorous process. [H.6]
        (Section 1R17.2.b.(1))


===B. Licensee-Identified Violations===
===Licensee-Identified Violations===


No violations were identified.
No violations were identified.
3


=REPORT DETAILS=
=REPORT DETAILS=


==REACTOR SAFETY==
==REACTOR SAFETY==
Cornerstone s: Initiating Events, Mitigating Systems, and Barrier Integrity 1R17 Evaluation s of Changes, Tests, and Experiments and Permanent Plant Modifications (71111.17 T)
Cornerstones: Initiating Events, Mitigating Systems, and Barrier Integrity {{a|1R17}}
==1R17 Evaluations of Changes, Tests, and Experiments and Permanent Plant Modifications==
{{IP sample|IP=IP 71111.17T}}
===.1 Evaluation of Changes, Tests, and Experiments===


===.1 Evaluation of Changes, Tests, and===
====a. Inspection Scope====
The inspectors reviewed 6 safety evaluations performed pursuant to Title 10, Code of Federal Regulations (CFR) 50.59 to determine if the evaluations were adequate and that prior U.S. Nuclear Regulatory Commission (NRC) approval was obtained as appropriate.


Experiments
The inspectors also reviewed 13 screenings and 2 applicability determinations where licensee personnel had determined that a 10 CFR 50.59 evaluation was not necessary.


====a. Inspection Scope====
The inspectors reviewed these documents to determine if:
The inspectors reviewed 6 safety evaluations performed pursuant to Title 10 , Code of Federal Regulations (CFR) 50.59 to determine if the evaluations were adequate and that prior U.S. Nuclear Regulatory Commission (NRC) approval was obtained as appropriate. The inspectors also reviewed 13 screenings and 2 applicability determinations where licensee personnel had determined that a 10 CFR 50.59 evaluation was not necessary. The inspectors reviewed these documents to determine if:
* the changes, tests, and 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 changes, tests, and 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, and experiments were correct and consistent with 10 CFR 50.59; and the design and licensing basis documentation was updated to reflect the change.
* the safety issue requiring the change, tests or experiment was resolved;
* the licensee conclusions for evaluations of changes, tests, and 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 6 samples of evaluations and 15 samples of screenings and/or applicability determinations as defined in Inspection Procedure (IP) 71111.17-04.
This inspection constituted 6 samples of evaluations and 15 samples of screenings and/or applicability determinations as defined in Inspection Procedure (IP) 71111.17-04.
Line 122: Line 106:


====a. Inspection Scope====
====a. Inspection Scope====
The inspectors reviewed six permanent plant modifications that had been installed in the plant during the last three years. This review included in
The inspectors reviewed six 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 control room; emergency diesel generators fuel oil storage tank level transmitter and indicator; fire door 519 A and 520A; and dampers CV 5325 B/C. The modifications were selected based upon risk significance, safety significance, and complexity. The inspectors reviewed the modifications selected to determine if:
-plant walkdowns for portions of the control room; emergency diesel generators fuel oil storage tank level transmitter and 4 indicator; fire door 519 A and 520A; and dampers CV 5325 B/C. The modifications were selected based upon risk significance, safety significance, and complexity. The inspectors reviewed the modifications selected to determine if:
* the supporting design and licensing basis documentation was updated;
the 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 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 constitut ed six permanent plant modification samples as defined in IP 71111.17-04.
This inspection constituted six permanent plant modification samples as defined in IP 71111.17-04.


====b. Findings====
====b. Findings====
: (1) Failure to Install and Control Reactor Coolant Pump (RCP) Seal Cavity Vent Flexible Hose s Per Design Basis Analysis Introduction
: (1) Failure to Install and Control Reactor Coolant Pump (RCP) Seal Cavity Vent Flexible     Hoses Per Design Basis Analysis
:  The inspectors identified a finding of very low safety significance (Green) and associated Non
 
-Cited Violation (NCV) of 10 CFR Part 50, Appendix B, Criterion III, "Design Control" for the licensee's failure to install and control the RCP seal cavity vent flexible hose s per the design basis analysis. Specifically, the licensee failed to correctly translate the design basis installation configuration and installation fatigue analysis in calculation SP I, "Pipe Stress
=====Introduction:=====
The inspectors identified a finding of very low safety significance (Green)and associated Non-Cited Violation (NCV) of 10 CFR Part 50, Appendix B, Criterion III, Design Control for the licensees failure to install and control the RCP seal cavity vent flexible hoses per the design basis analysis. Specifically, the licensee failed to correctly translate the design basis installation configuration and installation fatigue analysis in calculation SP-274-I, Pipe Stress


=====Analysis:=====
=====Analysis:=====
Reactor Coolant Pump 1 1 Seal Cavity Vent," into specifications, drawings, procedures, and instructions.
Reactor Coolant Pump 1-1-1 Seal Cavity     Vent, into specifications, drawings, procedures, and instructions.


Description
=====Description:=====
:  The Davis-Besse Reactor Coolant System (RCS) uses four RCPs. The RCPs are shaft
The Davis-Besse Reactor Coolant System (RCS) uses four RCPs. The RCPs are shaft-sealed with a seal cartridge assembly that consists of three mechanical face-type sealing stages. Each sealing stage has a cavity vent line. The vent lines are classified as American Society of Mechanical Engineers (ASME) Section III, Class 2 piping (3/4 inch pipe). The lines form part of the Reactor Coolant Pressure Boundary (RCPB) and are safety-related. Typically, piping that is part of the RCPB is required to be classified as ASME Class 1 piping per 10 CFR 50.55a(c)(1). However, the RCP seal cavity vent lines at Davis-Besse are classified as ASME Class 2 in accordance with 10 CFR 50.55a(c)(2) because in the event of a postulated failure of the lines, the reactor can be shut down and cooled down in an orderly manner assuming seal injection is maintained by the Makeup System.
-sealed with a seal cartridge assembly that consists of three mechanical face-type sealing stages. Each sealing stage has a cavity vent line. The vent lines are classified as American Society of Mechanical Engineers (ASME) Section III, Class 2 piping (3/4 inch pipe). The lines form part of the Reactor Coolant Pressure Boundary (RCPB) and are safety
-related. Typically, piping that is part of the RCPB is required to be classified as ASME Class 1 piping per 10 CFR 50.55a(c)(1). However, the RCP seal cavity vent lines at Davis
-Besse are classified as ASME Class 2 in accordance with 1 0 CFR 50.55a(c)(2) because in the event of a postulated failure of the lines
, the reactor can be shut down and cooled down in an orderly manner assuming seal injection is maintained by the Makeup System.


On August 6, 2012, and later on August 27, 2013, the licensee submitted Licensee Event Reports (LER s) to the NRC which described two different leaks from welds on the first stage seal cavity vent line for RCP 1-2. The LERs identified high
On August 6, 2012, and later on August 27, 2013, the licensee submitted Licensee Event Reports (LERs) to the NRC which described two different leaks from welds on the first stage seal cavity vent line for RCP 1-2. The LERs identified high-cycle fatigue as the failure mechanism for the welds. As part of the planned corrective actions to address the high-cycle fatigue, the licensee developed Engineering Change Package (ECP) 12-0785, Install Flexible Hose on RCP Seal 1st, 2nd, 3rd Stage Vents, Seal Injection and Controlled Bleedoff Lines, to replace a section of each of the RCP seal cavity vent lines with flexible hoses, among other things. The modification was implemented during the sites 18th refueling outage which began on February 1, 2014, and ended on May 6, 2014. A total of 12 flexible hoses were installed, three on each of the four RCPs. One end of each flexible hose was connected directly to its corresponding RCP, while the other end of the flexible hose was connected to the existing vent piping.
-cycle fatigue as 5 the failure mechanism for the welds. As part of the planned corrective actions to address the high
-cycle fatigue, the licensee developed Engineering Change Package (ECP) 12-0785, "Install Flexible Hose on RCP Seal 1st, 2nd, 3rd Stage Vents, Seal Injection and Controlled Bleedoff Lines
," to replace a section of each of the RCP seal cavity vent lines with flexible hoses, among other things. The modification was implemented during the site's 18th refueling outage which began on February 1, 2014, and ended on May 6, 2014. A total of 12 flexible hoses were installed, three on each of the four RCPs. One end of each flexible hose was connected directly to its corresponding RCP, while the other end of the flexible hose was connected to the existing vent piping.


The licensee procured the flexible hoses to meet the same design requirements as the original ASME Section III, Class 2 piping. The flexible hoses were designed, analyzed, and fabricated by a vendor. The vendor provided the licensee with the flexible hose design calculations, and these were incorporated into new pipe stress calculations for each RCP developed in support of ECP 12
The licensee procured the flexible hoses to meet the same design requirements as the original ASME Section III, Class 2 piping. The flexible hoses were designed, analyzed, and fabricated by a vendor. The vendor provided the licensee with the flexible hose design calculations, and these were incorporated into new pipe stress calculations for each RCP developed in support of ECP 12-0785. Through a review of the pipe stress calculation related to RCP 1-1-1, SP-274-I, Pipe Stress
-0785. Through a review of the pipe stress calculation related to RCP 1 1, SP-274-I, "Pipe Stress


=====Analysis:=====
=====Analysis:=====
Reactor Coolant Pump 1-1-1 Seal Cavity Vent," and discussions with the licensee, the inspectors identified two design control issues related to the flexible hoses.
Reactor Coolant Pump 1-1-1 Seal Cavity Vent, and discussions with the licensee, the inspectors identified two design control issues related to the flexible hoses.
 
The first issue identified by the inspectors was a non-conformance between the analyzed configuration of the flexible hoses by the vendor, and the installed configuration of the flexible hoses. The vendor analyzed the flexible hoses as being fixed (anchored) at the ends which attached to the existing piping, as documented in 5 of calculation SP-274-I. However, the inspectors identified that two-way restraints, instead of fixed restraints, were used to support the piping near those flexible hose ends. Fixed restraints would prevent displacement and rotation of the flexible hoses and piping in all directions. The installed two-way restraints allow the pipe and flexible hoses to displace in the axial direction, and also allow rotation in all directions.


The first issue identified by the inspectors was a non
Since the installed configuration of the flexible hoses and piping allows axial displacement, the inspectors questioned the qualification of the flexible hoses and piping under design basis loads, such as earthquakes. Specifically, the inspectors questioned whether the analysis for the flexible hoses had considered the additional seismic displacement loads resulting from the non-conformance of the installed configuration.
-conformance between the analyzed configuration of the flexible hoses by the vendor, and the installed configuration of the flexible hoses. The vendor analyzed the flexible hoses as being fixed (anchored) at the ends which attached to the existing piping, as documented in Attachment 5 of calculation SP I. However, the inspectors identified that two-way restraints, instead of fixed restraints, were used to support the piping near those flexible hose ends. Fixed restraints would prevent displacement and rotation of the flexible hoses and piping in all directions. The installed two
-way restraints allow the pipe and flexible hoses to displace in the axial direction, and also allow rotation in all directions. Since the installed configuration of the flexible hoses and piping allows axial displacement, the inspectors questioned the qualification of the flexible hoses and piping under design basis loads, such as earthquakes. Specifically, the inspectors questioned whether the analysis for the flexible hoses had considered the additional seismic displacement loads resulting from the non
-conformance of the installed configuration. The licensee's pipe stress analysis, including the vendor's flexible hose analysis, did not account for the non
-conformance.


This issue was entered into the licensee's Corrective Action Program (CAP) as CR-2014-17626, "2014 NRC MOD/50.59 Inspection
The licensees pipe stress analysis, including the vendors flexible hose analysis, did not account for the non-conformance.
- Evaluation of RCP Flex Hose Installed Configuration" and CR
-2014-17983, "2014 NRC MOD/50.59 Inspection
- Additional NRC Questions/Concerns on RCP Flex Hoses," which addressed the non
-conformance between the analyzed and installed configuration of the flexible hoses and the piping. The licensee determined that the non
-conformance could increase the seismic fatigue of the flexible hoses, based on discussions with the vendor and a review of the pipe stress calculations. However, because additional margin existed in the flexible hose fatigue analysis, the licensee concluded that the flexible hoses remained operable with reasonable expectation. The additional margin in the fatigue analysis had been incorporated into the thermal fatigue portion which evaluated a conservative RCS heat up and cool down fatigue life of 1550 cycles versus the actual RCS limit of 240 cycles. The licensee's planned corrective action for the issue at the conclusion of 6 the inspection was to obtain a revised calculation from the vendor and revise their pipe stress calculations to address the as
-built configuration of the flexible hoses and piping.


The second issue identified by the inspectors for the flexible hoses is related to the installation fatigue evaluated by the vendor in the flexible hose design analysis. The design calculations for the flexible hose s evaluated an installation fatigue of five cycles , as documented in Attachment 5 of calculation SP I. An installation cycle consists of one removal and re-installation of the flexible hose s. Therefore, in order for the design basis installation fatigue evaluation to remain valid, the licensee was required to control and limit the flexible hose installation cycles. During the inspection, the inspectors determined that the licensee did not have any controls in place to limit the installation cycles for the flexible hoses. The flexible hoses would, at a minimum, undergo an installation cycle during each RCP motor or seal replacement. The RCP seals are typically replaced every 8 years.
This issue was entered into the licensees Corrective Action Program (CAP) as CR-2014-17626, 2014 NRC MOD/50.59 Inspection - Evaluation of RCP Flex Hose Installed Configuration and CR-2014-17983, 2014 NRC MOD/50.59 Inspection -
Additional NRC Questions/Concerns on RCP Flex Hoses, which addressed the non-conformance between the analyzed and installed configuration of the flexible hoses and the piping. The licensee determined that the non-conformance could increase the seismic fatigue of the flexible hoses, based on discussions with the vendor and a review of the pipe stress calculations. However, because additional margin existed in the flexible hose fatigue analysis, the licensee concluded that the flexible hoses remained operable with reasonable expectation. The additional margin in the fatigue analysis had been incorporated into the thermal fatigue portion which evaluated a conservative RCS heat up and cool down fatigue life of 1550 cycles versus the actual RCS limit of 240 cycles. The licensees planned corrective action for the issue at the conclusion of the inspection was to obtain a revised calculation from the vendor and revise their pipe stress calculations to address the as-built configuration of the flexible hoses and piping.


Th is issue was entered into the licensee's Corrective Action Program (CAP) as CR-2014-17319, "2014 NRC MOD/50.59 Inspection  
The second issue identified by the inspectors for the flexible hoses is related to the installation fatigue evaluated by the vendor in the flexible hose design analysis. The design calculations for the flexible hoses evaluated an installation fatigue of five cycles, as documented in Attachment 5 of calculation SP-274-I. An installation cycle consists of one removal and re-installation of the flexible hoses. Therefore, in order for the design basis installation fatigue evaluation to remain valid, the licensee was required to control and limit the flexible hose installation cycles. During the inspection, the inspectors determined that the licensee did not have any controls in place to limit the installation cycles for the flexible hoses. The flexible hoses would, at a minimum, undergo an installation cycle during each RCP motor or seal replacement. The RCP seals are typically replaced every 8 years.
- Tracking of Installation Cycle Life for RCP Flex Hoses,"
 
which recognized that the flexible hoses had an installation limit of five cycles, and that the installation cycles were not being controlled because there was no tracking mechanism for these cycles. Currently, the flexible hoses have only accumulated one installation cycle because they were installed during the 2014 refueling outage and have not been removed since. Therefore, the flexible hoses have not exceeded the installation fatigue analysis limit, and there is reasonable expectation that all RCP seal cavity vent flexible hoses are operable. The licensee's planned corrective actions at the conclusion of the inspection for this issue were to either track the flexible hose installation cycles and replace the hoses prior to exceeding their installation cycle limit, or to revise the fatigue analysis to increase the number of installation cycles to the point where tracking of the cycles would not be required.
This issue was entered into the licensees Corrective Action Program (CAP) as CR-2014-17319, 2014 NRC MOD/50.59 Inspection - Tracking of Installation Cycle Life for RCP Flex Hoses, which recognized that the flexible hoses had an installation limit of five cycles, and that the installation cycles were not being controlled because there was no tracking mechanism for these cycles. Currently, the flexible hoses have only accumulated one installation cycle because they were installed during the 2014 refueling outage and have not been removed since. Therefore, the flexible hoses have not exceeded the installation fatigue analysis limit, and there is reasonable expectation that all RCP seal cavity vent flexible hoses are operable. The licensees planned corrective actions at the conclusion of the inspection for this issue were to either track the flexible hose installation cycles and replace the hoses prior to exceeding their installation cycle limit, or to revise the fatigue analysis to increase the number of installation cycles to the point where tracking of the cycles would not be required.


=====Analysis:=====
=====Analysis:=====
The inspectors determined the licensee's failure to install and control the RCP seal cavity vent flexible hose s in accordance with the design basis analysis was contrary to 10 CFR Part 50, Appendix B, Criterion III, "Design Control," and was a performance deficiency. The performance deficiency was determined to be more than minor because , if left uncorrected
The inspectors determined the licensees failure to install and control the RCP seal cavity vent flexible hoses in accordance with the design basis analysis was contrary to 10 CFR Part 50, Appendix B, Criterion III, Design Control, and was a performance deficiency. The performance deficiency was determined to be more than minor because, if left uncorrected, it would have the potential to lead to a more significant safety concern. Specifically, the failure to install and control the flexible hoses in accordance with the design basis analysis could lead to failure of the hoses due to operation beyond their analyzed limits.
, it would have the potential to lead to a more significant safety concern. Specifically, the failure to install and control the flexible hose s in accordance with the design basis analysis could lead to failure of the hose s due to operation beyond their analyzed limit s.


The inspectors determined the finding could be evaluated using the Significance Determination Process (S DP) in accordance with IMC 0609, "Significance Determination Process," Attachment 0609.04, "Initial Characterization of Findings," issued on June 19, 2012. Specifically, the inspectors used IMC 0609 Appendix A "SDP for Findings At
The inspectors determined the finding could be evaluated using the Significance Determination Process (SDP) in accordance with IMC 0609, Significance Determination Process, Attachment 0609.04, Initial Characterization of Findings, issued on June 19, 2012. Specifically, the inspectors used IMC 0609 Appendix A SDP for Findings At-Power, issued June 19, 2012, Exhibit 1, Initiating Events Screening Questions, to screen the finding. The Initiating Events Screening Questions were used instead of the Barrier Integrity Screening Questions because RCS Boundary issues are evaluated under the Initiating Events cornerstone unless they are related to pressurized thermal shock, as discussed in Section 5.0 of IMC 0609 Appendix A. The finding screened as of very low safety significance (Green) because the inspectors answered No to all of the screening questions in Subsection A, LOCA Initiators, of Exhibit 1.
-Power," issued June 19, 2012, Exhibit 1, "Initiating Events Screening Questions ," to screen the finding. The Initiating Events Screening Questions were used instead of the Barrier Integrity Screening Questions because RCS Boundary issues are evaluated under the Initiating Events cornerstone unless they are related to pressurized thermal shock, as discussed in Section 5.0 of IMC 0609 Appendix A. The finding screened as of very low safety significance (Green) because the inspectors answered "No" to all of the screening questions in Subsection A, "LOCA Initiators," of Exhibit 1.


Specifically, the finding could not result in exceeding the RCS leak rate for a small Loss of Coolant Accident (LOCA) after a reasonable assessment of degradation, and it could 7 not have likely affected other systems used to mitigate a LOCA resulting in a total loss of their function after a reasonable assessment of degradation. As discussed above, the flexible hoses remained operable and are not expected to fail due to the lack of design controls. In addition, the size of the lines is such that in the event of a postulated failure of a single line the reactor can be shut down and cooled down in an orderly manner assuming seal injection is maintained by the Makeup System.
Specifically, the finding could not result in exceeding the RCS leak rate for a small Loss of Coolant Accident (LOCA) after a reasonable assessment of degradation, and it could not have likely affected other systems used to mitigate a LOCA resulting in a total loss of their function after a reasonable assessment of degradation. As discussed above, the flexible hoses remained operable and are not expected to fail due to the lack of design controls. In addition, the size of the lines is such that in the event of a postulated failure of a single line the reactor can be shut down and cooled down in an orderly manner assuming seal injection is maintained by the Makeup System.


The inspectors determined this finding had an associated cross
The inspectors determined this finding had an associated cross-cutting aspect, Design Margins (H.6), in the Human Performance cross-cutting area. This corresponds to the apparent cause identified by the inspectors for the finding, the licensees failure to operate and maintain equipment within design margins, to carefully guard and change margins only through a systematic and rigorous process, and to place special attention to maintaining fission product barriers, defense-in-depth, and safety-related equipment.
-cutting aspect, Design Margins (H.6), in the Human Performance cross
-cutting area. This corresponds to the apparent cause identified by the inspectors for the finding, the licensee's failure to operate and maintain equipment within design margins, to carefully guard and change margins only through a systematic and rigorous process, and to place special attention to maintaining fission product barriers, defense
-in-depth, and safety
-related equipment. Specifically, the licensee did not carefully guard and change the RCP seal cavity vent lines, which form part of the RCS fission product barrier, through a systematic a nd rigorous process. This led to the failure to adequately install and control the flexible hoses in accordance with the design basis analysis.  [H.6]
Enforcement
:  10 CFR Part 50, Appendix B, Criterion III, "Design Control" requires, in part, that measures be established to assure that applicable regulatory requirements and the design basis are correctly translated into specifications, drawings, procedures, and instructions.


Contrary to the above, since about May 6, 2014, the licensee failed to assure that applicable regulatory requirements and the design basis were correctly translated into specifications, drawings, procedures, and instructions. Specifically, the licensee failed to install and control the RCP seal cavity vent flexible hoses per the design basis analysis in calculation SP I, "Pipe Stress
Specifically, the licensee did not carefully guard and change the RCP seal cavity vent lines, which form part of the RCS fission product barrier, through a systematic and rigorous process. This led to the failure to adequately install and control the flexible hoses in accordance with the design basis analysis. [H.6]
 
=====Enforcement:=====
10 CFR Part 50, Appendix B, Criterion III, Design Control requires, in part, that measures be established to assure that applicable regulatory requirements and the design basis are correctly translated into specifications, drawings, procedures, and instructions.
 
Contrary to the above, since about May 6, 2014, the licensee failed to assure that applicable regulatory requirements and the design basis were correctly translated into specifications, drawings, procedures, and instructions. Specifically, the licensee failed to install and control the RCP seal cavity vent flexible hoses per the design basis analysis in calculation SP-274-I, Pipe Stress


=====Analysis:=====
=====Analysis:=====
Reactor Coolant Pump 1 1 Seal Cavity Vent."  The licensee's immediate corrective actions taken to restore compliance were:
Reactor Coolant Pump 1-1-1 Seal Cavity Vent.
 
The licensees immediate corrective actions taken to restore compliance were:
: (1) to verify through discussions with the vendor that the flexible hoses would be expected to perform as designed under design basis loads, and
: (1) to verify through discussions with the vendor that the flexible hoses would be expected to perform as designed under design basis loads, and
: (2) to verify that the flexible hoses had not exceeded the installation fatigue analysis limit.
: (2) to verify that the flexible hoses had not exceeded the installation fatigue analysis limit.


Because this violation was of very
Because this violation was of very-low safety significance and was entered into the licensees CAP as CR-2014-17626, 2014 NRC MOD/50.59 Inspection - Evaluation of RCP Flex Hose Installed Configuration, CR-2014-17983, 2014 NRC MOD/50.59 Inspection - Additional NRC Questions/Concerns on RCP Flex Hoses, and CR-2014-17319, 2014 NRC MOD/50.59 Inspection - Tracking of Installation Cycle Life for RCP Flex Hoses, this violation is being treated as an NCV, consistent with Section 2.3.2 of the NRC Enforcement Policy. [NCV 05000346/2014007-01, Failure to Install and Control RCP Seal Cavity Vent Flexible Hoses Per Design Basis Analysis]
-low safety significance and was entered into the licensee's CAP as CR-2014-17626, "2014 NRC MOD/50.59 Inspection  
- Evaluation of RCP Flex Hose Installed Configuration," CR
-2014-17983, "2014 NRC MOD/50.59 Inspection  
- Additional NRC Questions/Concerns on RCP Flex Hoses,"
and CR-2014-17319, "2014 NRC MOD/50.59 Inspection  
- Tracking of Installation Cycle Life for RCP Flex Hoses," this violation is being treated as an NCV, consistent with Section 2.3.2 of the NRC Enforcement Policy. [NCV 05000346/2014007
-01 , "Failure to Install and Control RCP Seal Cavity Vent Flexible Hose s Per Design Basis Analysis"]  


==OTHER ACTIVITIES==
==OTHER ACTIVITIES (OA)==
(OA)
{{a|4OA2}}
{{a|4OA2}}
==4OA2 Problem Identification and Resolution==
==4OA2 Problem Identification and Resolution==
Line 217: Line 181:


====a. Inspection Scope====
====a. Inspection Scope====
The inspectors reviewed several 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 p lant modifications and evaluations of changes, tests, and experiments. 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.
The inspectors reviewed several corrective action process documents that identified or were related to 10 CFR 50.59 evaluations and permanent plant modifications. The inspectors reviewed these documents to evaluate the effectiveness of corrective actions related to permanent plant modifications and evaluations of changes, tests, and experiments. In addition, corrective action documents written on issues identified during the inspection were reviewed to verify adequate problem identification and incorporation of the problems into the corrective action system. The specific corrective action documents that were sampled and reviewed by the inspectors are listed in the attachment to this report.


====b. Findings====
====b. Findings====
: (1) Failure to Use Most Limiting 4160 Volts Alternating Current (VAC) Bus Voltage in Desig n Calculations Introduction
: (1) Failure to Use Most Limiting 4160 Volts Alternating Current (VAC) Bus Voltage in Design        Calculations
:  The inspectors identified an unresolved item (URI) for the licensee's failure to perform an analysis demonstrating that at the degraded voltage relay (DVR) set point specified in Technical Specifications (TS s) adequate voltage would be available to safety-related equipment to start and run during a design basis accident.
 
Description
:  On September 5, 2014, the licensee was informed by Schulz Electric that High Pressure Injection (HPI) Pump 1
-2 Motor will not meet all Purchase Order Requirements.
 
Specifically, the supplied motor may not be capable of accelerating the driven load to normal operating speed within 6.5 seconds at a minimum starting voltage of 70 percent. The licensee issued a condition report and initiated Prompt Operability Determination (POD) 2014
-13985, "HPI Pump 1-2 Motor" to evaluate the impact of the condition on the ability of the HPI Pump 1-2 Motor to perform its design function.
 
The licensee concluded that HPI 1-2 Pump was operable but nonconforming and placed a note to operators that in certain electrical plant alignments the HPI 1-2 Pump should be declared inoperable. On October 8, 2014, the licensee initiated CR
- 2014-15452, "Inconsistency in the Treatment of the Plant Restrictions in Operability Determinations".
 
The corrective actions documented the review of POD 2014-13985 to determine if the recommended actions/instructions to operators should have been considered compensatory actions and therefore screened in the licensee's 50.59 process.
 
The licensee concluded that the instructions to operators only provided information and therefore not considered a compensatory action.


During the review of POD 2014
=====Introduction:=====
-13985 , the inspectors were concerned that the licensee did not adequate ly address the operability of HPI Pump 1-2 Motor with respect to the DVR set point TS value of 3712 VAC.
The inspectors identified an unresolved item (URI) for the licensees failure to perform an analysis demonstrating that at the degraded voltage relay (DVR) set point specified in Technical Specifications (TSs) adequate voltage would be available to safety-related equipment to start and run during a design basis accident.


The inspectors reviewed NRC Regulatory Issue Summary (RIS) 2011-12, Revision 1, "Adequacy of Station Electrical Distribution System Voltages"; dated December 29, 2011, which was issued to clarify the NRC staff's 9 technical position on existing regulatory requirements.
=====Description:=====
On September 5, 2014, the licensee was informed by Schulz Electric that High Pressure Injection (HPI) Pump 1-2 Motor will not meet all Purchase Order Requirements. Specifically, the supplied motor may not be capable of accelerating the driven load to normal operating speed within 6.5 seconds at a minimum starting voltage of 70 percent. The licensee issued a condition report and initiated Prompt Operability Determination (POD) 2014-13985, HPI Pump 1-2 Motor to evaluate the impact of the condition on the ability of the HPI Pump 1-2 Motor to perform its design function. The licensee concluded that HPI 1-2 Pump was operable but nonconforming and placed a note to operators that in certain electrical plant alignments the HPI 1-2 Pump should be declared inoperable.


Specifically, the RIS clarifies voltage studies necessary for DVR setting bases.
On October 8, 2014, the licensee initiated CR- 2014-15452, Inconsistency in the Treatment of the Plant Restrictions in Operability Determinations. The corrective actions documented the review of POD 2014-13985 to determine if the recommended actions/instructions to operators should have been considered compensatory actions and therefore screened in the licensees 50.59 process. The licensee concluded that the instructions to operators only provided information and therefore not considered a compensatory action.


The RIS sta tes, in part, "the limiting voltage at the bus monitored by the DVR can be calculated in terms of the voltage at the terminals of the most limiting safety
During the review of POD 2014-13985, the inspectors were concerned that the licensee did not adequately address the operability of HPI Pump 1-2 Motor with respect to the DVR set point TS value of 3712 VAC. The inspectors reviewed NRC Regulatory Issue Summary (RIS) 2011-12, Revision 1, Adequacy of Station Electrical Distribution System Voltages; dated December 29, 2011, which was issued to clarify the NRC staffs technical position on existing regulatory requirements. Specifically, the RIS clarifies voltage studies necessary for DVR setting bases. The RIS states, in part, the limiting voltage at the bus monitored by the DVR can be calculated in terms of the voltage at the terminals of the most limiting safety-related component in the plant in all required operating conditions (such as starting and running). In addition, the RIS discusses that calculations of voltages at the terminals of all safety-related equipment with the voltage at the DVR monitored bus at the DVR dropout setting would ensure adequate voltage (starting and running) to all safety-related equipment.
-related component in the plant in all required operating conditions (such as starting and running)"
. In addition
, the RIS discusses that calculations of voltages at the terminals of all safety
-related equipment with the voltage at the DVR monitored bus at the DVR dropout setting would ensure adequate voltage (starting and running) to all safety
-related equipment.


The inspectors requested the licensee's DVR set point analysis as described in RIS 2011-12. The licensee had an analysis that shows all of the safety
The inspectors requested the licensees DVR set point analysis as described in RIS 2011-12. The licensee had an analysis that shows all of the safety-related loads would be able to run at steady-state at the TS DVR set point during design basis accidents. However, they do not have an analysis which shows all safety-related loads would start and run at the TS DVR set point.
-related loads would be able to run at steady
-state at the TS DVR set point during design basis accident s. However, they do not have an analysis which shows all safety
-related load s would start and run at the TS DVR set point
. The licensee's position is that if a design basis accident were to occur and the 4160 VAC Bus (Bus C1/D1)was at the Degraded Voltage set point of 3712 VAC , because the time delay is set at 7 seconds , they would divorce from the offsite power source and power would be supplied by the Emergency Diesel Generators (EDG s). Therefore, the licensee believes that they are not required to have an analysis which demonstrates that all required safety
-related loads can start and run at the DVR set point, as described in RIS 2011
-12, Revision 1.


Based on this information, the inspectors were concerned that the licensee does not have an analysis which demonstrates the safety
The licensees position is that if a design basis accident were to occur and the 4160 VAC Bus (Bus C1/D1) was at the Degraded Voltage set point of 3712 VAC, because the time delay is set at 7 seconds, they would divorce from the offsite power source and power would be supplied by the Emergency Diesel Generators (EDGs).
-related loads could start and run at the TS DVR set point.


The licensee captured the inspector's concerns in their CAP as CR 2014-17296 , "2014 50.59 Inspection:
Therefore, the licensee believes that they are not required to have an analysis which demonstrates that all required safety-related loads can start and run at the DVR set point, as described in RIS 2011-12, Revision 1.
Davis-Besse does not have an analysis to satisfy item 1 of RIS 2011
-12", dated November 19, 2014.


During the evaluation of CR 2014-17296 the licensee stated , "An analysis has been performed and demonstrates that all loads receive adequate voltage to start and perform their intended function with the exception of some Motor Operated Valves (MOVs). In order to ensure the MOVs have adequate voltage to perform their function, a minimum voltage of 4070 VAC should be maintained on either 4160 VAC Bus (C1/D1)
Based on this information, the inspectors were concerned that the licensee does not have an analysis which demonstrates the safety-related loads could start and run at the TS DVR set point. The licensee captured the inspectors concerns in their CAP as CR 2014-17296, 2014 50.59 Inspection: Davis-Besse does not have an analysis to satisfy item 1 of RIS 2011-12, dated November 19, 2014. During the evaluation of CR 2014-17296 the licensee stated, An analysis has been performed and demonstrates that all loads receive adequate voltage to start and perform their intended function with the exception of some Motor Operated Valves (MOVs). In order to ensure the MOVs have adequate voltage to perform their function, a minimum voltage of 4070 VAC should be maintained on either 4160 VAC Bus (C1/D1). Since the minimum voltage of 4070 VAC is higher than the previous voltage of 3900 VAC, the licensee has added a compensatory action to monitor Bus C1/ D1 at the higher voltage. This issue is unresolved pending consultation with Nuclear Reactor Regulation to determine if the licensee is required to demonstrate that safety-related loads can start and run at the DVR TS set point.
."  Since the minimum voltage of 4070 VAC is higher than the previous voltage of 3900 VAC
, the licensee has added a compensatory action to monitor Bus C1/ D1 at the higher voltage
. This issue is unresolved pending consultation with Nuclear Reactor Regulation to determine if the licensee is required to demonstrate that safety
-related loads can start and run at the DVR TS set point.


{{a|4OA6}}
{{a|4OA6}}
Line 277: Line 209:
===.1 Interim Meeting Summary===
===.1 Interim Meeting Summary===


On November 20, 2014, the inspectors presented the preliminary inspection results to Mr. K. Byrd and other members of the licensee staff.
On November 20, 2014, the inspectors presented the preliminary inspection results to Mr. K. Byrd and other members of the licensee staff. The licensee acknowledged the issues presented. The inspectors confirmed that none of the potential report input discussed was considered proprietary. The inspectors had outstanding questions that required additional review and a follow-up exit meeting.


The licensee acknowledged the issues presented.
===.2 Exit Meeting Summary===


The inspectors confirmed that none of the potential report input discussed was considered proprietary.
On January 9, 2015, the inspectors presented the inspection results to Mr. R. Lieb 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.


The inspectors had outstanding questions that required additional review and a follow
ATTACHMENT:
-up exit meeting.


===.2 Exit Meeting===
=SUPPLEMENTAL INFORMATION=


Summary  O n January 9, 2015
==KEY POINTS OF CONTACT==
, the inspector s presented the inspection results to Mr. R. Lieb and other members of the licensee staff. The licensee personnel acknowledged the inspection results presented and did not identify any proprietary content. The inspectors confirmed that all proprietary material reviewed during the inspection was returned to the licensee staff.


ATTACHMENT: 
Licensee
 
=SUPPLEMENTAL INFORMATION=
 
SUPPLEMENTAL INFORMATION
KEY POINTS OF CONTAC
T Licensee  
: [[contact::R. Lieb]], Site Vice President
: [[contact::R. Lieb]], Site Vice President
: [[contact::J. Hook]], Design Engineering Manager
: [[contact::J. Hook]], Design Engineering Manager
Line 303: Line 227:
: [[contact::G. Wolf]], Regulatory Compliance Supervisor
: [[contact::G. Wolf]], Regulatory Compliance Supervisor
: [[contact::V. Wadsworth]], Regulatory Compliance Specialist
: [[contact::V. Wadsworth]], Regulatory Compliance Specialist
: [[contact::U.S. Nuclear Regulatory Commission R. Daley]], Chief, Engineering Branch 3, DRS
U.S. Nuclear Regulatory Commission
: [[contact::R. Daley]], Chief, Engineering Branch 3, DRS
: [[contact::D. Kimble]], Senior Resident Inspector
: [[contact::D. Kimble]], Senior Resident Inspector
LIST OF ITEMS OPENED, CLOSED AND DISCUSS
ED Opened 05000346/20
00 7-01 NCV Failure to
Install and Control RCP Seal Cavity Vent Flexible Hose s Per Design Basis
Analysis (Section 1R17.2.b.(1))
05000346/20
00 7-02 URI Failure to Use Worst Case 4160 VAC Bus Voltage in Design Calculations (Section 4OA2.1.B(1)
) Closed 05000346/20
00 7-01 NCV Failure to Install and Control RCP Seal Cavity Vent Flexible Hose s Per Design Basis
Analysis (Section 1R17.2.b.(1))
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.
CFR 50.59 EVALUATIONS
Number Description or Title
Date or Revision
11-01921 Davis-Besse Spent Fuel Pool Criticality Analysis Revised Code Bias Uncertainty
Revision 0
13-00037 Replacement of Electro Hydraulic (EHC) and Turbine Supervisory Instrumentation (TSI) Systems
Revision 0
08-05743 Replacement of the Unit Load Demand (ULD) Function of the Integrated Control System (ICS)
Revision 0
11-02793 Auxiliary Building HELB Pressure Analysis Using GOTHIC 7.0 Revision 0
11-01972 Control Room Radiation Dosed Following a Maximum Hypothetical Accident
Revision 0
11-01623 Control Room Dose Due to the Fuel Handling Accident Inside Containment
Revision 0
CFR 50.59 SCREENINGS
Number Description or Title
Date or Revision
11-01921 Davis-Besse Spent Fuel Pool Criticality Analysis Revised Code Bias Uncertainty
Revision 0
2-02429 Install Freeze Seal on RCP 1
-2 Seal First Stage Vent Line
Revision 0
2-02751 Design Specification for the Installation of ASME Section III and ANSI B31.1 Piping
Revision 0
14-01385 RPS, SFAS, SFRCS, Trip or SG Tube Rupture Revision 0
13-00459 Replace Kaman Radiation Monitor
Revision 2
13-04584 Revised Tech Spec Bases 3.3.17, Action F.1
Revision 1
14-00366 Loss of AC Bus Power Source
Revision 0
14-01385 RPS, SFAS, SFRCS Trip or SG Tube Rupture
Revision 0
14-02390 Perform Hayes Reclosing Setting Changes
Revision 0
10-03148 Update Cable Jacketing Description in USAR & FHAR
Revision 0
10-03009 Change Thermal Overload Htr Element for MP35
Revision 0
10-02037 Update USAR to reflect TS Amendment 240 & Revise Vendor Drawing M
-519-50 under ECP 10
-0798 Revision 1
11-01958 Improve Performance Margin in CREATCS Air Cooled Condenser S61
-1, 2 Revision 4
11-00711 Equivalent Engineering Change 02
-0707 to Replace Three (3) Train 1 and Three (3) Train 2 Battery Chargers Revision 4
10-05651 Replacement of Pilot Operated Relief Valve RC2A
Revision 0


CALCULATIONS
==LIST OF ITEMS==
Number Description or Title
Date or Revision
C-NF-062.02-027 Davis-Besse Spent Fuel Pool Criticality An
alysis Revised Code Bias Uncertainty
Revision 0
HI-2002359 Criticality Analysis for Storage Racks in the Spent Fuel Pool of the Davis
-Besse Nuclear Power Station Revision 1
SP-274-I Pipe Stress Analysis:
Reactor Coolant Pump 1
-1-1 Seal Cavity Vent
Revision 0
WCAP-16518-P Beaver Valley Unit 2 Spent Fuel Pool Criticality Analysis Revision 2
C-ICE-026.02-003 Emergency Diesel Generator Fuel Oil Storage Tank Level
July 24, 2014
C-CSS-028.01-024 Seismic Calculation For Evaluation Of Air Cooled Condenser Refrigerant Pipe Routing
C-ME-028.01-011 CREVS Capacity Test
Revision 2
C-CSS-028.01-025 Seismic Evaluation for CREATCS Pressure Gage Assembly  CORRECTIVE ACTION PROGRAM DOCUMENTS
INITIATED DURING INSPECTION
Number Description or Title
Date or Revision 2014-17296 2014 50.59 Inspection:
Davis-Besse Does Not Have An Analysis to Satisfy Item
of RIS 2011
-12 November 19, 2014
2014-16611 2014 NRC MOD/ 50.59
- Procedure Not Followed For Editorial Changes
to 50.59 RAD and Eval
November 04, 2014
2014-17258 2014 NRC MOD/ 50.59 Inspection
- Freeze Seal in RCS Pressure Boundary
November 18, 2014
2014-17317 2014 NRC MOD/ 50.59 Inspection
- Vibration Not Addressed for Impact on RCP Flex hoses
November 19, 2014
2014-17321 2014 NRC MOD/ 50.59 Inspection
- Application of Code Cases for RCP Flex Piping
November 19, 2014
2014-17319 2014 NRC MOD/ 50.59 Inspection
- Tracking of Installation Cycle Life for RCP Flex Hoses
November 19, 2014
2014-17353 2014 NRC MOD/ 50.59 Inspection
- Status of Document Not Clearly Communicated
November 20, 2014
2014-17352 2014 NRC MOD/ 50.59 Inspection
- Documenting PM Revision in DIE Response
November 20, 2014
2014-17626 2014 NRC MOD/50.59 Inspection
- Evaluation of RCP Flex Hose Installed Configuration
November 26, 2014
2014-17983 2014 NRC MOD/50.59 Inspection
- Additional NRC Questions/Concerns on RCP Flex Hoses
December 5, 2014


CORRECTIVE ACTION PROGRAM DOCUMENTS REVIEWED
===OPENED, CLOSED AND DISCUSSED===
Number Description or Title Date or Revision
CR-2014-13985 HPI Pump 2 Motor Does Not Meet PO Requirements
September 9, 2014
CR-2014-14406 MU Pump 1 Motor Does Not Meet PO Requirement
September 15, 2014 CR-2014-15134 Part 21 Notification
- Shulz Electric Motors
October1, 2014
CR-2010- 74188 Cables Found with PVC Jackets
CR-2010-77047 Thermal OL Heaters for MP
-35  CR-2012-12992 Switchgear Room Ventilation During HELB
August 23, 2013
CR-2014-15452 Inconsistency
in the Treatment of Plant Restrictions in Operability Determinations October 08, 2014
DRAWINGS Number Description or Title
Date or Revision
E-1 SH.1 A.C. Electrical System One Line Diagram
E-2 SH.1 25KV & 13.8KV Metering and Relaying One Line Diagram 15 E-2 SH.2 25KV & 13.8KV Metering and Relaying One Line Diagram 8 E-3 4.16KV Metering and Relaying One Line Diagram
E-4 SH-1 "E" Buses 480V Unit Substations One Line Diagram
E-4 SH-2 "F" Buses 480V Unit Substations One Line Diagram
Revision 46 E-52B SH.5C Elementary Wiring Diagrams Reactor Cooling System HP INJ PMP 1
-2 Revision 3 E-52B SH.5D Elementary Wiring Diagrams Reactor Cooling System HP INJ PMP 1
-2 Revision 2 E-34B SH.13 4.16 KV FD BRKRS Bus C1(D1) Tripping & Lockout Relays & Synchro Check Relays
Revision 12 OS-056 SH 1 Operational Schematic 345KV System
Revision 17 OS-035  Operational Schematic Auxiliary Building Non
- Radioactive HVAC Systems
Revision 28
& 30  MODIFICATIONS
Number Description or Title
Date or Revision
2-0416 RCP 1-2 Seal Cavity Vent Freeze Seal Installation
Revision 0
2-0785 Install Flexible Hose on RCP Seal 1
st , 2 nd , 3 rd Stage Vents, Seal Injection and Controlled Bleedoff Lines
Revision 2
14-0262-001 Rescale of EDG Fuel Oil Storage Tank Level Transmitter and Indicator LT/L14891
Revision 3
10-0298 Corrections to RSM and SAP FLOC Database
Revision 0
10-0798 Update UFSAR for TS Amendment 240 Granting Fuel Transfer Tube Blind Flange Testing Exception
Revision 0
11-0265 Performance Improvements to CREATCS
Revision 2
2-0632 Closing Fire Doors 519 A and 520 A and Disabling Revision 2


MODIFICATIONS
===Opened===
Number Description or Title
: 05000346/2014007-01        NCV  Failure to Install and Control RCP Seal Cavity Vent Flexible Hoses Per Design Basis Analysis (Section 1R17.2.b.(1))
Date or Revision
: 05000346/2014007-02        URI  Failure to Use Worst Case 4160 VAC Bus Voltage in Design Calculations (Section 4OA2.1.B(1))
Dampers CV 5325B and CV 5325C
OPERABILITY EVALUATIONS
Number Description or Title
Date or Revision
POD 2014-13985 HPI Pump Motor 2 Prompt Operability Determination
Revision 0
OTHER DOCUMENTS
Number Description or Title
Date or Revision
SN-SA-2014-0560 2014 Full Self
-Assessment
October 31, 2014
PROCEDURES
Number Description or Title
Date or Revision
DB-MM-09012 RCP Seal Removal and Replacement
Revision 12
DB-OP-02000 RPS, SFAS, SFRCS Trip, or SG Tube Rupture
Revision 26 and 27
EN-DP-01107 Performing ASME Section XI Reconciliations
Revision 3
NOP-CC-5703 FirstEnergy Nuclear Operating Company (FENOC) ASME Section XI Repair/Replacement (R/R) Program Revision 2
DB-OP-02521 Loss of AC Bus Power Source, Attachment 1
Revision 22
DB-SC-03070 Emergency Diesel Generator 1 Monthly Test
Revision 35
Db-OP-06505 Control Room Emergency Ventilation System
DB-SS-03710 Functional Test for Control Room Emergency Ventilation System Train 1
DB-SS-03711 Functional Test for Control Room Emergency Ventilation System Train 2


LIST OF ACRONYMS USE
===Closed===
D ADAMS Agencywide Documents Access and Management System
: 05000346/2014007-01        NCV  Failure to Install and Control RCP Seal Cavity Vent Flexible Hoses Per Design Basis Analysis (Section 1R17.2.b.(1))
ASME American Society of Mechanical Engineers
CAP Corrective Action Program
CFR Code of Federal Regulations
DRS Division of Reactor Safety
DVR Degraded Voltage Relay
EDG Emergency Diesel Generator
ECP Engineering Change Package
FENOC First Energy Nuclear Operation Company
HPI High Pressure Injection
IMC Inspection Manual Chapter
IN Information Notice IP Inspection Procedure
LER Licensee Event Report
LOCA Loss of Coolant Accident
MOV Motor Operated Valve
NCV Non-Cited Violation
NEI Nuclear Energy Institute
NRC U.S. Nuclear Regulatory Commission
PARS Public Available Records System
POD Prompt Operability Determination
RCP Reactor Coolant Pump
RCPB Reactor Coolant Pressure Boundary
RCS Reactor Coolant System
RIS Regulatory Issue Summary
SDP Significance Determination Process
SP Surveillance Procedure
TS Technical Specification
URI Unresolved Item
VAC Volts Alternating Current
: [[contact::R. Lieb -2- In accordance with Title 10 of the Code of Federal Regulations (10 CFR) 2.390]], "Public Inspections, Exemptions, Requests for Withholding," 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's Public Document Room or from the Publicly Available Records (PARS) component of the NRC's Agencywide Documents Access and Management 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/
Robert
: [[contact::C. Daley]], Chief
Engineering Branch 3
Division of Reactor Safety
Docket Nos.
50-346; 72-017 License No. NPF-3 Enclosure:
Inspection Report 0500
0346/2014007  w/Attachment:  Supplemental Information
cc w/encl:
Distribution via LISTSERV  DISTRIBUTION
: See next page


ADAMS Accession Number ML15050
==LIST OF DOCUMENTS REVIEWED==
A 150  Publicly Available
Non-Publicly Available
Sensitive  Non-Sensitive To receive a copy of this document, indicate in the concurrence box "C" = Copy without attach/encl "E" = Copy with attach/encl  "N" = No copy
OFFICE RIII  RIII        NAME JGilliam:cl
RDaley    DATE 02/13/15 02/17/15    OFFICIAL RECORD COPY


Letter to Mr. Raymond Lieb from Mr. Robert
C.Daley dated February 17, 2015
SUBJECT:  DAVIS-BESSE NUCLEAR POWER STATION EVALUATIONS OF CHANGES, TESTS, AND EXPERIMENTS AND PERMANENT PLANT MODIFICATIONS BASELINE INSPECTION REPORT 05000346/2014
007  DISTRIBUTION w/encl
: Kimyata MorganButler
RidsNrrDorlLpl3
-2 Resource
RidsNrrPMDavisBesse Resource
RidsNrrDirsIrib Resource
Cynthia Pederson
Darrell Roberts
Eric Duncan
Allan Barker
Carole Ariano
Linda Linn
DRPIII DRSIII Jim Clay Carmen Olteanu
ROPreports.Resource@nrc.gov
}}
}}

Latest revision as of 05:50, 20 December 2019

IR 05000346/2014007; on 11/03/2014 - 01/09/2015; Davis-Besse Nuclear Power Station; Evaluations of Changes, Tests, and Experiments and Permanent Plant Modifications
ML15050A150
Person / Time
Site: Davis Besse, Trojan  Cleveland Electric icon.png
Issue date: 02/17/2015
From: Robert Daley
Engineering Branch 3
To: Lieb R
FirstEnergy Nuclear Operating Co
References
IR 2014007
Download: ML15050A150 (20)
v • d • e


Text

UNITED STATES ary 17, 2015

SUBJECT:

DAVIS-BESSE NUCLEAR POWER STATION EVALUATIONS OF CHANGES, TESTS, AND EXPERIMENTS AND PERMANENT PLANT MODIFICATIONS BASELINE INSPECTION REPORT 05000346/2014007

Dear Mr. Lieb:

On January 9, 2015, the U.S. Nuclear Regulatory Commission (NRC) completed an Evaluations of Changes, Tests, and Experiments and Permanent Plant Modifications Inspection at your Davis-Besse Nuclear Power Station. The enclosed inspection report documents the inspection results which were discussed on January 9, 2015, with you 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.

One NRC-identified finding of very-low safety significance (Green) was identified during this inspection. This finding was determined to involve a violation of NRC requirements. However, because of the very-low safety significance and because the issue was entered into your corrective action program, the NRC is treating the issue as a Non-Cited Violation (NCV) in accordance with Section 2.3.2 of the NRC Enforcement Policy.

If you contest the violation or significance of the 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 copies to the Regional Administrator, U.S. Nuclear Regulatory Commission - Region III; the Director, Office of Enforcement, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001; and the Resident Inspector at Davis-Besse Nuclear Power 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 Davis-Besse Nuclear Power Station. In accordance with Title 10 of the Code of Federal Regulations (10 CFR) 2.390, Public Inspections, Exemptions, Requests for Withholding, 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 NRCs Public Document Room or from the Publicly Available Records (PARS)

component of the NRC's Agencywide Documents Access and Management 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/

Robert C. Daley, Chief Engineering Branch 3 Division of Reactor Safety Docket Nos. 50-346;72-017 License No. NPF-3

Enclosure:

Inspection Report 05000346/2014007 w/Attachment: Supplemental Information

REGION III==

Docket Nos: 50-346;72-017 License No: NPF-3 Report No: 05000346/2014007 Licensee: FirstEnergy Nuclear Operating Company (FENOC)

Facility: Davis-Besse Nuclear Power Station Location: Oak Harbor, OH Dates: November 3, 2014 - January 09, 2015 Inspectors: Jasmine Gilliam, Acting Senior Reactor Inspector (Lead)

Loyd Cain, Senior Resident Inspector (Vogtle)

Ijaz Hafeez, Reactor Inspector Lionel Rodriguez, Reactor Inspector (Observer)

Approved by: Robert C. Daley, Chief Engineering Branch 3 Division of Reactor Safety Enclosure

SUMMARY OF FINDINGS

IR 05000346/2014007; 11/03/2014 - 01/09/2015; Davis-Besse Nuclear Power Station;

Evaluations of Changes, Tests, and Experiments and Permanent Plant Modifications.

This report covers a two-week announced baseline inspection on evaluations of changes, tests, and experiments and permanent plant modifications. The inspection was conducted by Region III based engineering inspectors. One finding of very-low safety significance was identified by the inspectors. The finding was considered a Non-Cited Violation (NCV) of U.S. Nuclear Regulatory Commission (NRC) regulations. The significance of most findings is indicated by their color (i.e., greater than Green, or Green, White, Yellow, Red) using Inspection Manual Chapter (IMC) 0609, Significance Determination Process (SDP). Cross-cutting aspects were determined using IMC 0310, Aspects 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. All violations of NRC requirements are dispositioned in accordance with the NRCs Enforcement Policy dated July 9, 2013. The NRCs program for overseeing the safe operation of commercial nuclear power reactors is described in NUREG-1649, Reactor Oversight Process, Revision 5, dated February 2014.

NRC-Identified

and Self-Revealed Findings

Cornerstone: Initiating Events

Green.

The inspectors identified a finding of very-low safety significance (Green) and associated Non-Cited Violation (NCV) of 10 CFR Part 50, Appendix B, Criterion III,

Design Control for the licensees failure to install and control the Reactor Coolant Pump (RCP) seal cavity vent flexible hoses per the design basis analysis. Specifically, the licensee failed to correctly translate the design basis installation configuration and installation fatigue analysis in calculation SP-274-I, Pipe Stress Analysis: Reactor Coolant Pump 1-1-1 Seal Cavity Vent, into specifications, drawings, procedures, and instructions. The licensee entered this finding into their Corrective Action Program (CAP) to review the lack of controls over the installation of the flexible hoses, but determined that the flexible hoses remained operable.

The performance deficiency was determined to be more than minor because, if left uncorrected, it would have the potential to lead to a more significant safety concern.

Specifically, the failure to install and control the flexible hoses in accordance with the design basis analysis could lead to failure of the hoses due to operation beyond their analyzed limits. The finding screened as of very-low safety significance (Green)because the finding could not result in exceeding the Reactor Coolant System (RCS)leak rate for a small Loss of Coolant Accident (LOCA) after a reasonable assessment of degradation, and it could not have likely affected other systems used to mitigate a LOCA resulting in a total loss of their function after a reasonable assessment of degradation.

The inspectors determined this finding had an associated cross-cutting aspect, Design Margins, in the Human Performance cross-cutting area. Specifically, the licensee did not carefully guard and change the RCP seal cavity vent lines, which form part of the RCS fission product barrier, through a systematic and rigorous process. [H.6]

(Section 1R17.2.b.(1))

Licensee-Identified Violations

No violations were identified.

REPORT DETAILS

REACTOR SAFETY

Cornerstones: Initiating Events, Mitigating Systems, and Barrier Integrity

1R17 Evaluations of Changes, Tests, and Experiments and Permanent Plant Modifications

Main article: IP 71111.17T

.1 Evaluation of Changes, Tests, and Experiments

a. Inspection Scope

The inspectors reviewed 6 safety evaluations performed pursuant to Title 10, Code of Federal Regulations (CFR) 50.59 to determine if the evaluations were adequate and that prior U.S. Nuclear Regulatory Commission (NRC) approval was obtained as appropriate.

The inspectors also reviewed 13 screenings and 2 applicability determinations where licensee personnel had determined that a 10 CFR 50.59 evaluation was not necessary.

The inspectors reviewed these documents to determine if:

  • the changes, tests, and 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, and 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 6 samples of evaluations and 15 samples of screenings and/or applicability determinations as defined in Inspection Procedure (IP) 71111.17-04.

b. Findings

No findings were identified.

.2 Permanent Plant Modifications

a. Inspection Scope

The inspectors reviewed six 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 control room; emergency diesel generators fuel oil storage tank level transmitter and indicator; fire door 519 A and 520A; and dampers CV 5325 B/C. The modifications were selected based upon risk significance, safety significance, and complexity. The inspectors reviewed the modifications selected to determine if:

  • the supporting design and licensing basis documentation was updated;
  • the changes were in accordance with the specified design requirements;
  • the procedures and training plans affected by the modification have been adequately updated;
  • the test documentation as required by the applicable test programs has been updated; and
  • post-modification testing adequately verified system operability and/or functionality.

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

This inspection constituted six permanent plant modification samples as defined in IP 71111.17-04.

b. Findings

(1) Failure to Install and Control Reactor Coolant Pump (RCP) Seal Cavity Vent Flexible Hoses Per Design Basis Analysis
Introduction:

The inspectors identified a finding of very low safety significance (Green)and associated Non-Cited Violation (NCV) of 10 CFR Part 50, Appendix B, Criterion III, Design Control for the licensees failure to install and control the RCP seal cavity vent flexible hoses per the design basis analysis. Specifically, the licensee failed to correctly translate the design basis installation configuration and installation fatigue analysis in calculation SP-274-I, Pipe Stress

Analysis:

Reactor Coolant Pump 1-1-1 Seal Cavity Vent, into specifications, drawings, procedures, and instructions.

Description:

The Davis-Besse Reactor Coolant System (RCS) uses four RCPs. The RCPs are shaft-sealed with a seal cartridge assembly that consists of three mechanical face-type sealing stages. Each sealing stage has a cavity vent line. The vent lines are classified as American Society of Mechanical Engineers (ASME)Section III, Class 2 piping (3/4 inch pipe). The lines form part of the Reactor Coolant Pressure Boundary (RCPB) and are safety-related. Typically, piping that is part of the RCPB is required to be classified as ASME Class 1 piping per 10 CFR 50.55a(c)(1). However, the RCP seal cavity vent lines at Davis-Besse are classified as ASME Class 2 in accordance with 10 CFR 50.55a(c)(2) because in the event of a postulated failure of the lines, the reactor can be shut down and cooled down in an orderly manner assuming seal injection is maintained by the Makeup System.

On August 6, 2012, and later on August 27, 2013, the licensee submitted Licensee Event Reports (LERs) to the NRC which described two different leaks from welds on the first stage seal cavity vent line for RCP 1-2. The LERs identified high-cycle fatigue as the failure mechanism for the welds. As part of the planned corrective actions to address the high-cycle fatigue, the licensee developed Engineering Change Package (ECP) 12-0785, Install Flexible Hose on RCP Seal 1st, 2nd, 3rd Stage Vents, Seal Injection and Controlled Bleedoff Lines, to replace a section of each of the RCP seal cavity vent lines with flexible hoses, among other things. The modification was implemented during the sites 18th refueling outage which began on February 1, 2014, and ended on May 6, 2014. A total of 12 flexible hoses were installed, three on each of the four RCPs. One end of each flexible hose was connected directly to its corresponding RCP, while the other end of the flexible hose was connected to the existing vent piping.

The licensee procured the flexible hoses to meet the same design requirements as the original ASME Section III, Class 2 piping. The flexible hoses were designed, analyzed, and fabricated by a vendor. The vendor provided the licensee with the flexible hose design calculations, and these were incorporated into new pipe stress calculations for each RCP developed in support of ECP 12-0785. Through a review of the pipe stress calculation related to RCP 1-1-1, SP-274-I, Pipe Stress

Analysis:

Reactor Coolant Pump 1-1-1 Seal Cavity Vent, and discussions with the licensee, the inspectors identified two design control issues related to the flexible hoses.

The first issue identified by the inspectors was a non-conformance between the analyzed configuration of the flexible hoses by the vendor, and the installed configuration of the flexible hoses. The vendor analyzed the flexible hoses as being fixed (anchored) at the ends which attached to the existing piping, as documented in 5 of calculation SP-274-I. However, the inspectors identified that two-way restraints, instead of fixed restraints, were used to support the piping near those flexible hose ends. Fixed restraints would prevent displacement and rotation of the flexible hoses and piping in all directions. The installed two-way restraints allow the pipe and flexible hoses to displace in the axial direction, and also allow rotation in all directions.

Since the installed configuration of the flexible hoses and piping allows axial displacement, the inspectors questioned the qualification of the flexible hoses and piping under design basis loads, such as earthquakes. Specifically, the inspectors questioned whether the analysis for the flexible hoses had considered the additional seismic displacement loads resulting from the non-conformance of the installed configuration.

The licensees pipe stress analysis, including the vendors flexible hose analysis, did not account for the non-conformance.

This issue was entered into the licensees Corrective Action Program (CAP) as CR-2014-17626, 2014 NRC MOD/50.59 Inspection - Evaluation of RCP Flex Hose Installed Configuration and CR-2014-17983, 2014 NRC MOD/50.59 Inspection -

Additional NRC Questions/Concerns on RCP Flex Hoses, which addressed the non-conformance between the analyzed and installed configuration of the flexible hoses and the piping. The licensee determined that the non-conformance could increase the seismic fatigue of the flexible hoses, based on discussions with the vendor and a review of the pipe stress calculations. However, because additional margin existed in the flexible hose fatigue analysis, the licensee concluded that the flexible hoses remained operable with reasonable expectation. The additional margin in the fatigue analysis had been incorporated into the thermal fatigue portion which evaluated a conservative RCS heat up and cool down fatigue life of 1550 cycles versus the actual RCS limit of 240 cycles. The licensees planned corrective action for the issue at the conclusion of the inspection was to obtain a revised calculation from the vendor and revise their pipe stress calculations to address the as-built configuration of the flexible hoses and piping.

The second issue identified by the inspectors for the flexible hoses is related to the installation fatigue evaluated by the vendor in the flexible hose design analysis. The design calculations for the flexible hoses evaluated an installation fatigue of five cycles, as documented in Attachment 5 of calculation SP-274-I. An installation cycle consists of one removal and re-installation of the flexible hoses. Therefore, in order for the design basis installation fatigue evaluation to remain valid, the licensee was required to control and limit the flexible hose installation cycles. During the inspection, the inspectors determined that the licensee did not have any controls in place to limit the installation cycles for the flexible hoses. The flexible hoses would, at a minimum, undergo an installation cycle during each RCP motor or seal replacement. The RCP seals are typically replaced every 8 years.

This issue was entered into the licensees Corrective Action Program (CAP) as CR-2014-17319, 2014 NRC MOD/50.59 Inspection - Tracking of Installation Cycle Life for RCP Flex Hoses, which recognized that the flexible hoses had an installation limit of five cycles, and that the installation cycles were not being controlled because there was no tracking mechanism for these cycles. Currently, the flexible hoses have only accumulated one installation cycle because they were installed during the 2014 refueling outage and have not been removed since. Therefore, the flexible hoses have not exceeded the installation fatigue analysis limit, and there is reasonable expectation that all RCP seal cavity vent flexible hoses are operable. The licensees planned corrective actions at the conclusion of the inspection for this issue were to either track the flexible hose installation cycles and replace the hoses prior to exceeding their installation cycle limit, or to revise the fatigue analysis to increase the number of installation cycles to the point where tracking of the cycles would not be required.

Analysis:

The inspectors determined the licensees failure to install and control the RCP seal cavity vent flexible hoses in accordance with the design basis analysis was contrary to 10 CFR Part 50, Appendix B, Criterion III, Design Control, and was a performance deficiency. The performance deficiency was determined to be more than minor because, if left uncorrected, it would have the potential to lead to a more significant safety concern. Specifically, the failure to install and control the flexible hoses in accordance with the design basis analysis could lead to failure of the hoses due to operation beyond their analyzed limits.

The inspectors determined the finding could be evaluated using the Significance Determination Process (SDP) in accordance with IMC 0609, Significance Determination Process, Attachment 0609.04, Initial Characterization of Findings, issued on June 19, 2012. Specifically, the inspectors used IMC 0609 Appendix A SDP for Findings At-Power, issued June 19, 2012, Exhibit 1, Initiating Events Screening Questions, to screen the finding. The Initiating Events Screening Questions were used instead of the Barrier Integrity Screening Questions because RCS Boundary issues are evaluated under the Initiating Events cornerstone unless they are related to pressurized thermal shock, as discussed in Section 5.0 of IMC 0609 Appendix A. The finding screened as of very low safety significance (Green) because the inspectors answered No to all of the screening questions in Subsection A, LOCA Initiators, of Exhibit 1.

Specifically, the finding could not result in exceeding the RCS leak rate for a small Loss of Coolant Accident (LOCA) after a reasonable assessment of degradation, and it could not have likely affected other systems used to mitigate a LOCA resulting in a total loss of their function after a reasonable assessment of degradation. As discussed above, the flexible hoses remained operable and are not expected to fail due to the lack of design controls. In addition, the size of the lines is such that in the event of a postulated failure of a single line the reactor can be shut down and cooled down in an orderly manner assuming seal injection is maintained by the Makeup System.

The inspectors determined this finding had an associated cross-cutting aspect, Design Margins (H.6), in the Human Performance cross-cutting area. This corresponds to the apparent cause identified by the inspectors for the finding, the licensees failure to operate and maintain equipment within design margins, to carefully guard and change margins only through a systematic and rigorous process, and to place special attention to maintaining fission product barriers, defense-in-depth, and safety-related equipment.

Specifically, the licensee did not carefully guard and change the RCP seal cavity vent lines, which form part of the RCS fission product barrier, through a systematic and rigorous process. This led to the failure to adequately install and control the flexible hoses in accordance with the design basis analysis. [H.6]

Enforcement:

10 CFR Part 50, Appendix B, Criterion III, Design Control requires, in part, that measures be established to assure that applicable regulatory requirements and the design basis are correctly translated into specifications, drawings, procedures, and instructions.

Contrary to the above, since about May 6, 2014, the licensee failed to assure that applicable regulatory requirements and the design basis were correctly translated into specifications, drawings, procedures, and instructions. Specifically, the licensee failed to install and control the RCP seal cavity vent flexible hoses per the design basis analysis in calculation SP-274-I, Pipe Stress

Analysis:

Reactor Coolant Pump 1-1-1 Seal Cavity Vent.

The licensees immediate corrective actions taken to restore compliance were:

(1) to verify through discussions with the vendor that the flexible hoses would be expected to perform as designed under design basis loads, and
(2) to verify that the flexible hoses had not exceeded the installation fatigue analysis limit.

Because this violation was of very-low safety significance and was entered into the licensees CAP as CR-2014-17626, 2014 NRC MOD/50.59 Inspection - Evaluation of RCP Flex Hose Installed Configuration, CR-2014-17983, 2014 NRC MOD/50.59 Inspection - Additional NRC Questions/Concerns on RCP Flex Hoses, and CR-2014-17319, 2014 NRC MOD/50.59 Inspection - Tracking of Installation Cycle Life for RCP Flex Hoses, this violation is being treated as an NCV, consistent with Section 2.3.2 of the NRC Enforcement Policy. [NCV 05000346/2014007-01, Failure to Install and Control RCP Seal Cavity Vent Flexible Hoses Per Design Basis Analysis]

OTHER ACTIVITIES (OA)

4OA2 Problem Identification and Resolution

.1 Routine Review of Condition Reports

a. Inspection Scope

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

b. Findings

(1) Failure to Use Most Limiting 4160 Volts Alternating Current (VAC) Bus Voltage in Design Calculations
Introduction:

The inspectors identified an unresolved item (URI) for the licensees failure to perform an analysis demonstrating that at the degraded voltage relay (DVR) set point specified in Technical Specifications (TSs) adequate voltage would be available to safety-related equipment to start and run during a design basis accident.

Description:

On September 5, 2014, the licensee was informed by Schulz Electric that High Pressure Injection (HPI) Pump 1-2 Motor will not meet all Purchase Order Requirements. Specifically, the supplied motor may not be capable of accelerating the driven load to normal operating speed within 6.5 seconds at a minimum starting voltage of 70 percent. The licensee issued a condition report and initiated Prompt Operability Determination (POD) 2014-13985, HPI Pump 1-2 Motor to evaluate the impact of the condition on the ability of the HPI Pump 1-2 Motor to perform its design function. The licensee concluded that HPI 1-2 Pump was operable but nonconforming and placed a note to operators that in certain electrical plant alignments the HPI 1-2 Pump should be declared inoperable.

On October 8, 2014, the licensee initiated CR- 2014-15452, Inconsistency in the Treatment of the Plant Restrictions in Operability Determinations. The corrective actions documented the review of POD 2014-13985 to determine if the recommended actions/instructions to operators should have been considered compensatory actions and therefore screened in the licensees 50.59 process. The licensee concluded that the instructions to operators only provided information and therefore not considered a compensatory action.

During the review of POD 2014-13985, the inspectors were concerned that the licensee did not adequately address the operability of HPI Pump 1-2 Motor with respect to the DVR set point TS value of 3712 VAC. The inspectors reviewed NRC Regulatory Issue Summary (RIS) 2011-12, Revision 1, Adequacy of Station Electrical Distribution System Voltages; dated December 29, 2011, which was issued to clarify the NRC staffs technical position on existing regulatory requirements. Specifically, the RIS clarifies voltage studies necessary for DVR setting bases. The RIS states, in part, the limiting voltage at the bus monitored by the DVR can be calculated in terms of the voltage at the terminals of the most limiting safety-related component in the plant in all required operating conditions (such as starting and running). In addition, the RIS discusses that calculations of voltages at the terminals of all safety-related equipment with the voltage at the DVR monitored bus at the DVR dropout setting would ensure adequate voltage (starting and running) to all safety-related equipment.

The inspectors requested the licensees DVR set point analysis as described in RIS 2011-12. The licensee had an analysis that shows all of the safety-related loads would be able to run at steady-state at the TS DVR set point during design basis accidents. However, they do not have an analysis which shows all safety-related loads would start and run at the TS DVR set point.

The licensees position is that if a design basis accident were to occur and the 4160 VAC Bus (Bus C1/D1) was at the Degraded Voltage set point of 3712 VAC, because the time delay is set at 7 seconds, they would divorce from the offsite power source and power would be supplied by the Emergency Diesel Generators (EDGs).

Therefore, the licensee believes that they are not required to have an analysis which demonstrates that all required safety-related loads can start and run at the DVR set point, as described in RIS 2011-12, Revision 1.

Based on this information, the inspectors were concerned that the licensee does not have an analysis which demonstrates the safety-related loads could start and run at the TS DVR set point. The licensee captured the inspectors concerns in their CAP as CR 2014-17296, 2014 50.59 Inspection: Davis-Besse does not have an analysis to satisfy item 1 of RIS 2011-12, dated November 19, 2014. During the evaluation of CR 2014-17296 the licensee stated, An analysis has been performed and demonstrates that all loads receive adequate voltage to start and perform their intended function with the exception of some Motor Operated Valves (MOVs). In order to ensure the MOVs have adequate voltage to perform their function, a minimum voltage of 4070 VAC should be maintained on either 4160 VAC Bus (C1/D1). Since the minimum voltage of 4070 VAC is higher than the previous voltage of 3900 VAC, the licensee has added a compensatory action to monitor Bus C1/ D1 at the higher voltage. This issue is unresolved pending consultation with Nuclear Reactor Regulation to determine if the licensee is required to demonstrate that safety-related loads can start and run at the DVR TS set point.

4OA6 Management Meetings

.1 Interim Meeting Summary

On November 20, 2014, the inspectors presented the preliminary inspection results to Mr. K. Byrd and other members of the licensee staff. The licensee acknowledged the issues presented. The inspectors confirmed that none of the potential report input discussed was considered proprietary. The inspectors had outstanding questions that required additional review and a follow-up exit meeting.

.2 Exit Meeting Summary

On January 9, 2015, the inspectors presented the inspection results to Mr. R. Lieb and other members of the licensee staff. The licensee personnel acknowledged the inspection results presented and did not identify any proprietary content. The inspectors confirmed that all proprietary material reviewed during the inspection was returned to the licensee staff.

ATTACHMENT:

SUPPLEMENTAL INFORMATION

KEY POINTS OF CONTACT

Licensee

R. Lieb, Site Vice President
J. Hook, Design Engineering Manager
G. Michael, Design Engineering Supervisor
G. Wolf, Regulatory Compliance Supervisor
V. Wadsworth, Regulatory Compliance Specialist

U.S. Nuclear Regulatory Commission

R. Daley, Chief, Engineering Branch 3, DRS
D. Kimble, Senior Resident Inspector

LIST OF ITEMS

OPENED, CLOSED AND DISCUSSED

Opened

05000346/2014007-01 NCV Failure to Install and Control RCP Seal Cavity Vent Flexible Hoses Per Design Basis Analysis (Section 1R17.2.b.(1))
05000346/2014007-02 URI Failure to Use Worst Case 4160 VAC Bus Voltage in Design Calculations (Section 4OA2.1.B(1))

Closed

05000346/2014007-01 NCV Failure to Install and Control RCP Seal Cavity Vent Flexible Hoses Per Design Basis Analysis (Section 1R17.2.b.(1))

LIST OF DOCUMENTS REVIEWED

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