Text

IR 05000285/2014009 and Notice of Violation, July 7 Through September 12, 2014, Fort Calhoun, NRC Confirmatory Action Letter Follow Up Inspection and Problem Identification and Resolution
	ML14261A455
Person / Time
Site:	Fort Calhoun
Issue date:	09/18/2014
From:	Anton Vegel; Division of Reactor Safety II
To:	Cortopassi L; Omaha Public Power District
References
	EA-14-151 IR 2014009
	Download: ML14261A455 (141)
	v • d • e

UNITED STATES

NUCLEAR REGULATORY COMMISSION

REGION IV

1600 E LAMAR BLVD

ARLINGTON, TX 76011-4511

September 18, 2014

EA-14-151

Louis P. Cortopassi, Vice President

and Chief Nuclear Officer

Omaha Public Power District

Fort Calhoun Station FC-2-4

P.O. Box 550

Fort Calhoun, NE 68023-0550

SUBJECT: FORT CALHOUN STATION - NRC CONFIRMATORY ACTION LETTER

FOLLOW UP INSPECTION AND PROBLEM IDENTIFICATION AND

RESOLUTION INSPECTION REPORT NUMBER 05000285/2014009 AND

NOTICE OF VIOLATION

Dear Mr. Cortopassi:

On September 12, 2014, the U.S. Nuclear Regulatory Commission (NRC) completed a

Confirmatory Action Letter follow-up and Problem Identification and Resolution (PI&R) team

inspection at the Fort Calhoun Station (FCS). On September 12, 2014, the NRC inspection

team discussed the results of this inspection with you and other members of your staff. The

inspection team documented the results of this inspection in the enclosed inspection report.

The inspection focused on assessing activities related to the effectiveness of the FCS corrective

action program (CAP) and Omaha Public Power Districts (OPPDs) implementation of the

commitments described in Confirmatory Action Letter (CAL) EA-13-243, issued December 17,

2013 (ML13351A4231). CAL EA-13-243 confirmed the OPPDs commitments to ensure the

improvements realized during the extended outage remain in place and performance continues

to improve at the facility. Specifically, the NRC reviewed the CAL items associated with

10 Performance Improvement Integrated Matrix Action Plans characterized as the Key Drivers

for Achieving and Sustaining Excellence.

In performing the portion of the inspection associated with PI&R, the inspection team assessed

OPPDs threshold for identifying problems, implementation of the process for prioritizing and

evaluating problems, as well as the effectiveness of corrective actions identified and

implemented to resolve the problems. The team also evaluated the effectiveness of other

processes used to identify issues for resolution. These included the use of audits and self-

assessments, and incorporation of lessons learned from industry operating experience, into

station programs, processes, and procedures.

1

Designation in parentheses refers to an Agencywide Documents Access and Management System

(ADAMS) accession number. Unless otherwise noted, documents referenced in this letter are publicly

available using the accession number in ADAMS.

Based on the inspection results, the team concluded that FCS maintained a CAP in which

individuals generally identified issues at an appropriately low threshold. Of concern, the team

identified a number of deficiencies involving OPPDs ability to effectively and consistently

evaluate and resolve problems as discussed in more detail below, and in the enclosed

inspection report.

A number of the deficiencies that were identified by the inspectors involved inadequate

evaluations of degraded or non-conforming conditions that were entered into the CAP. Several

examples involved the failure to make an immediate determination of operability because your

staff failed to recognize that a degraded or non-conforming condition existed. Additional

examples involved operability evaluations that lacked adequate technical justification as to why

the affected system, structure, or component would perform its specified safety function as

described in licensing and design basis documentation. The team concluded weakness exists

in this area and that OPPDs corrective actions, to date, have not been effective in ensuring that

problems are consistently evaluated in a thorough and effective manner, with the appropriate

technical rigor that supports the operability conclusion.

These findings and observations are similar to previous NRC inspection results identified during

the last NRC PI&R team inspection completed in June 2013, documented in NRC Inspection

Report 05000285/2013008. Many of the issues, identified both prior to and following restart,

reflect a poor understanding and use of design basis information. We do note that you are

taking action to address this challenge by a long-term licensee commitment to conduct a design

basis reconstitution through 2018. Based on the results of this inspection, the NRC

understands OPPD is conducting evaluations to determine the cause of the performance

problems affecting effective implementation of the corrective action program. The NRC is

interested in understanding the status of these evaluations and corrective actions during an

upcoming public meeting in Omaha, Nebraska on September 25, 2014. Additionally, the NRC

will be conducting future inspections focused on these areas of concern.

The NRC determined that your staff appropriately evaluated industry operating experience for

relevance to the facility and entered applicable items into the CAP. Audits and self-

assessments were generally thorough and complete; however, the team identified instances

where the corrective actions to address the findings identified from these audits and self-

assessments were not always complete or timely. The NRC determined that your stations

management maintains a safety-conscious work environment in which your employees are

willing to raise nuclear safety concerns through at least one of the several means available.

The NRC determined that, with the exception of the PI&R key area and the Operability

Determination element of the Program key area, OPPD is adequately implementing the CAL

items. The NRC determined that five key CAL areas were adequately completed and are

considered closed. These five key areas are:

Organizational Effectiveness, Safety Culture, and Safety Conscious Work Environment

Site Operational Focus

Procedures

Nuclear Oversight

Transition to the Exelon Nuclear Management Model and Integration into the Exelon

Nuclear Fleet

With respect to the PI&R key area, the license had completed most of the items. Based on the

NRCs independent assessment results that identified the concerns previously discussed, this

key area will remain open pending further NRC inspection. The details of the NRCs

assessment of the Confirmatory Action Letter key performance areas are discussed in

Section 4OA4 of the enclosed report.

Finally, there were also a number of deficiencies identified by the inspectors that involved

inadequate resolution of problems. Of particular concern, the inspectors reviewed OPPDs

actions to resolve 36 previously issued NRC non-cited violations, documented in various

inspection reports in 2013, and identified five examples where OPPD failed to adequately

address the issues. Several examples were noted where no actions were either planned or

implemented to resolve the findings. The inspectors noted that a self-assessment, performed

by both OPPD and Exelon individuals prior to the team inspection, also identified this concern,

however, incomplete CAP implementation resulted in deficiencies that were not entered into the

process, and, subsequently, the particular non-cited violations were not adequately addressed.

As a result, the NRC has determined that one Severity Level IV violation of NRC requirements

occurred and four violations associated with findings of very low safety significance (Green)

occurred. The NRC evaluated these violations in accordance Section 2.3.2.a of the NRC

Enforcement Policy, which appears on the NRCs Web site at http://www.nrc.gov/about-

nrc/regulatory/enforcement/enforce-pol.html.

The NRC determined that these violations did not meet the criteria to be treated as non-cited

violations, and therefore will be cited in the enclosed Notice of Violation (Notice). These

violations are being cited because FCS failed to restore compliance (or demonstrate objective

evidence of plans to restore compliance) within a reasonable time after the violations were first

identified in NRC Inspection Report 05000285/2013008. You are required to respond to this

letter and should follow the instructions specified in the enclosed Notices of Violation when

preparing your response. If you have additional information that you believe the NRC should

consider, you may provide it in your response to the Notice. The NRCs review of your

response to the Notices will also determine whether further enforcement action is necessary to

ensure compliance with regulatory requirements.

The enclosed report documents 14 additional findings of very low safety significance (Green).

All of these findings involved violations of NRC requirements; one of these violations was

determined to be Severity Level IV under the traditional enforcement process. Two additional

Severity Level IV violations with no associated finding are also documented in the enclosed

report. The NRC is treating these violations as non-cited violations consistent with

Section 2.3.2.a of the NRC Enforcement Policy.

Four licensee-identified violations are being documented in the enclosed report that were

determined to be of very low safety significance. The NRC is treating these violations as non-

cited violations consistent with Section 2.3.2.a of the NRC Enforcement Policy.

If you contest the violations or significance of these violations, 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, Region IV; the

Director, Office of Enforcement, U.S. Nuclear Regulatory Commission,

Washington, DC 20555-0001; and the NRC resident inspector at Fort Calhoun Station.

If you disagree with a cross-cutting aspect assignment 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 IV; and the NRC resident inspector at Fort

Calhoun Station.

In accordance with Title 10 of the Code of Federal Regulations (10 CFR) 2.390, Public

Inspections, Exemptions, Requests for Withholding, a copy of this letter, its enclosures, 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/

Anton Vegel, Director

Division of Reactor Safety

Docket No.: 50-285

License No.: DPR-40

Enclosures:

1. Notice of Violation EA -14-151

2. NRC Inspection Report 05000285/2014009

w/Attachments:

1. Supplemental Information

2. Notification of Inspection and Request for Information

cc w/ encl: Electronic Distribution

ML14261A455

SUNSI Review Non-Sensitive Publicly Available Keyword:

By: MHay Sensitive Non-Publicly Available

OFFICE SRI:DRP/D SRI:DRP/D SPE:DRP/D ADD/DRP OE DD:DRS

NAME GWarnick JGroom RHagar MHay RBrowder AVegel

SIGNATURE /RA Email/ /RA Email/ /RA Email/ /RA/ /RA Email/ /RA/

DATE 9/17/14 9/17/14 9/15/14 9/18/14 9/18/14 9/18/14

SUBJECT: FORT CALHOUN STATION - NRC CONFIRMATORY ACTION LETTER

FOLLOW UP INSPECTION AND PROBLEM IDENTIFICATION AND

RESOLUTION INSPECTION REPORT NUMBER 05000285/2014009 AND

NOTICE OF VIOLATION

DISTRIBUTION:

Regional Administrator (Marc.Dapas@nrc.gov)

Deputy Regional Administrator (Kriss.Kennedy@nrc.gov)

Acting DRP Director (Troy.Pruett@nrc.gov)

Acting DRP Deputy Director (Michael.Hay@nrc.gov)

DRS Director (Anton.Vegel@nrc.gov)

DRS Deputy Director (Jeff.Clark@nrc.gov)

Senior Resident Inspector (Max.Schneider@nrc.gov)

Acting Senior Project Engineer, DRP/F (Peter.Jayroe@nrc.gov)

Acting Senior Project Engineer, DRP/F (Nick.Taylor@nrc.gov)

FCS Administrative Assistant (Janise.Schwee@nrc.gov)

RIV Public Affairs Officer (Victor.Dricks@nrc.gov)

RIV Public Affairs Officer (Lara.Uselding@nrc.gov)

NRR Project Manager (Fred.Lyon@nrc.gov)

RIV Branch Chief, DRS/TSB (Geoffrey.Miller@nrc.gov)

RIV RITS Coordinator (Marisa.Herrera@nrc.gov)

RIV Regional Counsel (Karla.Fuller@nrc.gov)

Congressional Affairs Officer (Angel.Moreno@nrc.gov)

RidsOEMailCenter.Resource

OEWEB Resource (Sue.Bogle@nrc.gov)

Technical Support Assistant (Loretta.Williams@nrc.gov)

RIV/ETA: OEDO (John.Jandovitz@nrc.gov)

RIV RSLO (Bill.Maier@nrc.gov)

ACES (R4Enforcement.Resource@nrc.gov)

MC 0350 Panel Chairman (Anton.Vegel@nrc.gov)

MC 0350 Panel Vice Chairman (Louise.Lund@nrc.gov)

MC 0350 Panel Member (Michael.Balazik@nrc.gov)

MC 0350 Panel Member (Michael.Markley@nrc.gov)

ROPreports

NOTICE OF VIOLATION

Omaha Public Power District Docket No: 50-285

Fort Calhoun Station License No: DPR-40

EA-14-151

During an NRC Inspection conducted from June 23 through September 12, 2014, violations of

NRC requirements were identified. In accordance with the NRC Enforcement Policy, the

violations are listed below:

A. 10 CFR Part 50, Appendix B, Criterion III, Design Control, requires in part that

measures shall be established to assure that applicable regulatory requirements and the

design basis, as defined in §50.2, and as specified in the license application, for those

structures, systems, and components to which this appendix applies are correctly

translated into specifications, drawings, procedures, and instructions.

Contrary to the above, as of November 28, 2010, measures established by the licensee

did not assure that applicable regulatory requirements and design bases were correctly

translated into specifications, drawings, procedures, and instructions. Specifically, the

licensee failed to properly evaluate NRC Bulletin 88-04, Potential Safety-Related Pump

Loss, for strong pump, weak pump, interaction regarding auxiliary feedwater pumps

FW-6 and FW-10, which are considered safety-related pumps. The licensee's

evaluation documented in Calculation FC08310, Auxiliary Feedwater (AFW) Motor

Driven Pump FW-6 and Turbine Driven Pump FW-10 Performance and Runout

Evaluation, Revision 0, failed to consider pump-to-pump interaction that may result due

to pump discharge check valve leakage.

This violation is associated with a Green Significance Determination Process finding.

B. 10 CFR Part 50, Appendix B, Criterion III, Design Control, requires, in part, that

measures shall be established to assure that applicable regulatory requirements and the

design basis, as defined in §50.2, and as specified in the license application, for those

structures, systems, and components to which this appendix applies are correctly

translated into specifications, drawings, procedures, and instructions.

Licensee's procedure AOP-01, Acts of Nature, Revision 33, instructs operators to

secure the raw water pumps at an intake cell level of 976'9".

Contrary to the above, from initial plant operations to present, measures established by

the licensee failed to assure that applicable regulatory requirements and the design

basis for those components are correctly translated into specifications, drawings,

procedures, and instructions. Specifically, the licensee failed to ensure that raw water

cooling was provided down to the design basis low river level of 976'9" mean sea level.

The intake cell level in the licensee's procedure AOP-01, is not equivalent to mean sea

level. As a result, the licensee failed to ensure the associated specifications and

procedures support raw water pump operations, which are safety related pumps, to

support the plant's cooling systems.

This violation is associated with a Green Significance Determination Process finding.

C. 10 CFR Part 50.54(q)(2), Conditions of License, requires, in part, that a nuclear power

reactor licensee shall follow and maintain the effectiveness of an emergency plan that

E1-1 Enclosure 1

meets the requirements of Appendix E to Part 50 and the planning standards of 10 CFR

50.47(b).

10 CFR 50.47(b)(4), requires, in part, that a standard emergency classification and

action level scheme, is in use by the nuclear facility licensee.

Contrary to the above, as of May 14, 2009, the licensee failed to maintain the

effectiveness of the emergency plan, by not maintaining a standard emergency

classification and action level scheme. Specifically, the emergency action level scheme

was not maintained because emergency action level HA1, Natural or destructive

phenomena affecting the Protected Area, contained an inaccurate river level of 9739

mean sea level. The river level was inaccurate because the basis document, Procedure

TBD-EPIP-OSC-1H, Recognition Category H - Hazards and Other Conditions Affecting

Plant Safety, Revision 2, stated the emergency action level was based on the minimum

elevation of the raw water pump suction. Because the river level does not correspond to

intake cell level, then the river level would have to be at least 973'10" mean sea level to

provide an adequate suction for the raw water pumps.

This violation is associated with a Green Significance Determination Process finding.

D. 10 CFR Part 50, Appendix B, Criterion III, Design Control, requires, in part, that

measures shall be established to assure that applicable regulatory requirements and the

design basis, as defined in §50.2, and as specified in the license application, for those

structures, systems, and components to which this appendix applies are correctly

translated into specifications, drawings, procedures, and instructions.

Engineering Analysis FC-92-072, Diesel Generator Loading Transient Analysis Using

Paladin Design Base 4.0, Revision 7, discussed a frequency spectrum of 60.5 +/-0.3

hertz for the emergency diesel generators, which are safety-related components.

Licensee's Calculation FC08034, Diesel Fuel Usage During a Severe Flooding Event,

does not assume that the diesel generators were run at 60.8 hertz for the entire 7-day

mission time.

Contrary to the above, as of June 2011, measures established by the licensee failed to

assure that applicable regulatory requirements and the design basis for those

components are correctly translated into specifications, drawings, procedures, and

instructions. Specifically, the licensee's calculation for fuel consumption did not assume

that the diesel generators were run at 60.8 hertz, for the entire 7-day mission time. As a

result, the licensee failed to translate the worst-case design emergency diesel generator

frequency of 60.8 hertz, which could impact the consumption of fuel oil, into the

applicable design documentation.

This violation is associated with a Green Significance Determination Process finding.

E. 10 CFR Part 50.59(c)(2)(ii), Changes, Tests, and Experiments, requires, in part, that a

licensee shall obtain a license amendment prior to implementing a proposed change,

test, or experiment if the change, test, or experiment would result in more than a minimal

increase in the likelihood of occurrence of a malfunction of a structure, system, or

component important to safety previously evaluated in the final safety analysis report (as

updated).

E1-2

10 CFR 50.59(d)(1) requires, in part, that the licensee shall maintain records of changes

in the facility or procedures and that the records must include a written evaluation which

provides the bases for the determination that the change does not require a license

amendment.

Contrary to the above, as of June 2008, the licensee did not perform a written evaluation

for a design change that may have required NRC review and approval. Specifically, the

licensee did not evaluate a change that would permanently substitute manual actions for

an automatic action to add water and nitrogen gas to the component cooling water surge

tank, which is an Updated Safety Analysis Report described design function for the

component cooling water system.

This violation is associated with a Severity Level IV traditional enforcement violation.

Pursuant to the provisions of 10 CFR 2.201, Omaha Public Power District is hereby required to

submit a written statement or explanation to the U.S. Nuclear Regulatory Commission, ATTN:

Document Control Desk, Washington, DC 20555-0001 with a copy to the Regional

Administrator, Region IV, and a copy to the NRC Resident Inspector at Fort Calhoun Station

within 30 days of the date of the letter transmitting this Notice of Violation (Notice). This reply

should be clearly marked as a Reply to Notice of Violation; EA 14-151, and should include:

(1) the reason for the violation, or, if contested, the basis for disputing the violation or severity

level, (2) the corrective steps that have been taken and the results achieved, (3) the corrective

steps that will be taken to avoid further violations, and (4) the date when full compliance will be

achieved. Your response may reference or include previous docketed correspondence, if the

correspondence adequately addresses the required response. If an adequate reply is not

received within the time specified in this Notice, an order or a Demand for Information may be

issued as to why the license should not be modified, suspended, or revoked, or why such other

action as may be proper should not be taken. Where good cause is shown, consideration will

be given to extending the response time.

If you contest this enforcement action, you should also provide a copy of your response, with

the basis for your denial, to the Director, Office of Enforcement, United States Nuclear

Regulatory Commission, Washington, DC 20555-0001.

Because your response will be made available electronically for public inspection in the NRC

Public Document Room or from the NRCs document system (ADAMS), accessible from the

NRC website at www.nrc.gov/reading-rm/adams.html, to the extent possible, it should not

include any personal privacy, proprietary, or safeguards information so that it can be made

available to the public without redaction. If personal privacy or proprietary information is

necessary to provide an acceptable response, then please provide a bracketed copy of your

response that identifies the information that should be protected and a redacted copy of your

response that deletes such information. If you request withholding of such material, you must

specifically identify the portions of your response that you seek to have withheld and provide in

detail the bases for your claim of withholding (e.g., explain why the disclosure of information will

create an unwarranted invasion of personal privacy or provide the information required by 10

CFR 2.390(b) to support a request for withholding confidential commercial or financial

information).

Dated this 18th day of September, 2014

E1-3

U.S. NUCLEAR REGULATORY COMMISSION

REGION IV

Docket: 05000285

License: DPR-40

Report: 05000285/2014009

Licensee: Omaha Public Power District

Facility: Fort Calhoun Station

Location: 9610 Power Lane

Blair, NE 68008

Dates: July 7 through September 12, 2014

Team Lead: G. Warnick, Senior Resident Inspector

Inspectors: I. Anchondo, Reactor Inspector

E. Coffman, Resident Inspector

B. Davis, Senior Construction Inspector

W. Deschaine, Resident Inspector

J. Groom, Senior Resident Inspector, Assistant Team Leader

B. Hagar, Senior Project Engineer, Assistant Team Leader

C. Henderson, Resident Inspector

D. Holman, Senior Security Specialist

G. Khouri, Senior Construction Inspector

J. Mateychick, Senior Reactor Inspector

C. Smith, Project Engineer

M. Williams, Reactor Inspector

Approved By: Anton Vegel

Director, Division of Reactor Safety

E2-1 Enclosure 2

SUMMARY

IR 05000285/2014009; 07/07/2014 - 09/12/2014; Fort Calhoun Station; Problem Identification

and Resolution Inspection and Confirmatory Action Letter Follow-up Inspection.

The inspection activities described in this report were performed from July 7-25, 2014, by

13 inspectors from the NRCs Region IV and Region II offices. The report documents

14 findings of very low safety significance (Green). All of these findings involved violations of

NRC requirements; one of these violations was determined to be Severity Level IV under the

traditional enforcement process. Additionally, NRC inspectors documented two Severity

Level IV violations with no associated finding. Further, NRC inspectors documented one

Severity Level IV violation and four violations associated with findings of very low safety

significance (Green) that were evaluated in accordance Section 2.3.2.a of the NRC

Enforcement Policy. The NRC inspectors determined that these violations did not meet the

criteria to be treated as non-cited violations because the licensee did not restore compliance

within a reasonable time after previous non-cited violations were issued.

The significance of inspection findings is indicated by their color (Green, White, Yellow, or Red),

which is determined using Inspection Manual Chapter 0609, Significance Determination

Process. Their cross-cutting aspects are determined using Inspection Manual Chapter 0310,

Components Within the Cross-Cutting Areas. Violations of NRC requirements are

dispositioned in accordance with the NRC Enforcement Policy. The NRC's program for

overseeing the safe operation of commercial nuclear power reactors is described in

NUREG-1649, Reactor Oversight Process.

Assessment of Problem Identification and Resolution

The NRC team reviewed approximately 400 condition reports, work orders, engineering

evaluations, root and apparent cause evaluations, and other supporting documentation to

determine if problems were being properly identified, characterized, and entered into the

corrective action program for evaluation and resolution. The team also reviewed a sample of

system health reports, self-assessments, trending reports and metrics, and various other

documents related to the corrective action program.

Based on its inspection sample, the team concluded that the licensee maintained a corrective

action program in which individuals generally identified issues at an appropriately low threshold.

Once entered into the corrective action program, the NRC noted deficiencies in the licensees

ability to effectively evaluate and resolve issues. The results of this inspection closely mirror

inspection results from previous team inspections conducted in 2011 and 2013 documented in

NRC Inspection Reports 05000285/2011006 and 05000285/2013008, respectively, in that the

NRC identified significant weakness in the licensees ability to evaluate and resolve issues

entered into the corrective action program. In particular, the team noted that technical

evaluations performed by the licensee were in some cases incomplete or contained incorrect

conclusions. The team identified several instances where the licensee failed to perform

appropriate evaluations for equipment issues entered into the corrective action program.

Specific examples include inadequate operability evaluations, technically inaccurate calculations

(used as corrective actions), and evaluations that failed to consider the design and licensing

basis of the facility or all applicable regulatory requirements. The findings and observations

identified by the team revealed significant weakness in the evaluation area and cause concern

related to the licensees ability to implement this element of the corrective action program.

E2-2

The team concluded that the licensee did not consistently develop appropriate corrective

actions to address issues entered into the corrective action program. The team noted that while

the licensee was identifying and placing a large number of adverse conditions into the corrective

action process, the associated resolution of these issues was often incomplete, narrowly

focused, or untimely. The team identified multiple examples of untimely or ineffective corrective

actions to address conditions adverse to quality. Of particular concern, the team reviewed the

licensees corrective actions to address 36 previous NRC non-cited violations and identified

5 examples where the licensee failed to restore compliance within a reasonable time after the

previous NRC violations were issued. Three other examples identified by the team involved the

failure to adequately address the technical aspects of the violation. Additionally, the team

identified several examples where the corrective action to address complex regulatory or

technical issues were incomplete, narrowly focused, or untimely. The findings and observations

identified by the team revealed significant weakness in the area of resolution and cause concern

related to the licensees ability to implement this element of the corrective action program.

The licensee appropriately evaluated industry operating experience for relevance to the facility

and entered applicable items in the corrective action program. The licensee incorporated

industry and internal operating experience in its root cause and apparent cause evaluations.

The licensee performed effective and self-critical nuclear oversight audits and self-assessments.

However, the corrective actions to address the individual findings from these audits and self-

assessments were not always complete or timely.

The licensee maintained a safety-conscious work environment in which personnel were willing

to raise nuclear safety concerns without fear of retaliation.

Cornerstone: Mitigating Systems

Green. Multiple examples of a non-cited violation of 10 CFR Part 50, Appendix B,

Criterion V, Instructions, Procedures, and Drawings, was identified involving the failure to

follow Procedure OP-FC-108-115, Operability Determinations, Revision 0a. In each

example, the team identified that the licensee failed to make an immediate determination of

operability for a degraded or non-conforming condition or failed to make an immediate

determination of operability based on a detailed examination of the deficiency. The licensee

took immediate corrective actions to update the incomplete or inaccurate operability

determinations and entered the collective failures to follow station operability procedures

into their corrective action program as Condition Report 2014-09163.

This performance deficiency was more than minor, and therefore a finding, because it

affected the equipment performance attribute of the Mitigating Systems Cornerstone

objective of ensuring the reliability of systems that respond to initiating events. The NRC

performed an initial screening of the finding in accordance with NRC Manual Chapter

IMC 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-

Power. Using IMC 0609, Appendix A, Exhibit 2, Mitigating Systems Screening Questions,

dated July 1, 2012, this finding is of very low safety significance (Green) because it: (1) was

not a deficiency affecting the design or qualification of a mitigating system; (2) did not

represent a loss of system and/or function; (3) did not represent an actual loss of function of

a single train for greater than its technical specification allowed outage time; and (4) does

not represent an actual loss of function of one or more non-technical specification trains of

equipment designated as high safety-significant in accordance with the licensees

maintenance rule program for greater than 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . This finding has a cross-cutting aspect

in the area of human performance because the licensee failed to use decision-making

E2-3

practices that demonstrate that a proposed action is to be safe in order to proceed, rather

than unsafe in order to stop. Specifically, the licensee made non-conservative decisions

related to the impact of degraded or non-conforming conditions [H.14]. (Section 4OA2.5.b)

SLIV/Green. A non-cited violation of 10 CFR 50.59, Changes, Tests, and Experiments,

and associated non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions,

Procedures, and Drawings, was identified involving the failure to evaluate and implement

adequate compensatory measures for a degraded condition associated with raw water

pump AC-10C. Specifically, the licensees operability determination established a

compensatory measure to place pump AC-10C in pull-to-lock, contrary to the system single

failure analysis design criteria described in the Updated Safety Analysis Report. The

licensee entered this issue into its corrective action program as Condition

Reports 2014-09104 and 2014-08515 and performed an operability evaluation and

associated 10 CFR 50.59 evaluation that used an acceptable compensatory measure to

pump water from affected manholes prior to affecting the degraded power feeder cable for

raw water pump AC-10C.

The NRC evaluated this performance deficiency as both a reactor oversight process finding

and a traditional enforcement violation. The NRC performed an initial screening of the

finding in accordance with NRC Manual Chapter IMC 0609, Appendix A, The Significance

Determination Process (SDP) for Findings At-Power. Using IMC 0609, Appendix A,

Exhibit 2, Mitigating Systems Screening Questions, dated July 1, 2012, this finding is of

very low safety significance (Green) because it: (1) was not a deficiency affecting the

design or qualification of a mitigating system; (2) did not represent a loss of system and/or

function; (3) did not represent an actual loss of function of a single train for greater than its

technical specification allowed outage time; and (4) does not represent an actual loss of

function of one or more non-technical specification trains of equipment designated as high

safety-significant for greater than 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months in accordance with the licensees maintenance

rule program. This finding has a cross-cutting aspect in the area of problem identification

and resolution with an aspect of evaluation because the licensee failed to ensure that

resolutions address causes and extent of conditions commensurate with their safety

significance [P.2].

In addition, because this performance deficiency had the potential to impact the NRCs

ability to perform its regulatory function in that the failure to obtain a license amendment for

a change that could result in a malfunction of a structure, system or component with a

different result than previously evaluated in the Updated Safety Analysis Report is in

violation of 10 CFR 50.59(c)(2)(vi), the NRC also evaluated the violation using traditional

enforcement. Since this violation is associated with a Green reactor oversight process

violation, the traditional enforcement violation was determined to be a Severity Level IV

violation, consistent with the example in paragraph 6.1.d(2) of the NRC Enforcement Policy.

(Section 4OA2.5.c)

Green. A non-cited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control,

was identified involving the failure to implement appropriate design control measures

associated with a safety-related pipe stress calculation. Specifically, several unverified and

potentially non-conservative inputs were identified associated with Calculation FC07240

used to analyze stresses on a pipe reduction tee in the safety injection system. The

licensee entered this issue into the corrective action program as Condition

Report 2014-09098 and initiated action to update Calculation FC07240.

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This performance deficiency was more than minor, and therefore a finding, because it

affected the design control attribute of the Mitigating Systems Cornerstone objective to

ensure the availability, reliability, and capability of components that respond to initiating

events. The NRC performed an initial screening of the finding in accordance with NRC

Manual Chapter IMC 0609, Appendix A, The Significance Determination Process (SDP) for

Findings At-Power. Using IMC 0609, Appendix A, Exhibit 2, Mitigating Systems Screening

Questions, dated July 1, 2012, this finding is of very low safety significance (Green)

because it: (1) was not a deficiency affecting the design or qualification of a mitigating

system; (2) did not represent a loss of system and/or function; (3) did not represent an

actual loss of function of a single train for greater than its technical specification allowed

outage time; and (4) does not represent an actual loss of function of one or more non-

technical specification trains of equipment designated as high safety-significant in

accordance with the licensees maintenance rule program for greater than 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . This

finding has a cross-cutting aspect in the area of human performance in that the licensee

failed to apply the appropriate rigor when evaluating the overstressed pipe union tee [H.6].

(Section 4OA2.5.e)

Green. A self-revealing non-cited violation of 10 CFR Part 50, Appendix B, Criterion III,

Design Control, was identified involving the failure to maintain design control of the raw

water strainer AC-12B control panel AI-348. Specifically, the licensee failed to adequately

design control panel AI-348 to protect it from the effects of spraying and wetting as required

by the plants licensing and design basis. The licensee entered this issue into its corrective

action program as Condition Reports 2013-03301 and 2014-06974 and initiated action to

encase control panel AI-348 to protect it against the effects of spraying and wetting.

This performance deficiency was more than minor, and therefore a finding, because it was

associated with the equipment performance attribute of the Mitigating Systems Cornerstone

and affected the associated objective to ensure availability, reliability, and capability of

systems that respond to initiating events to prevent undesirable consequences. Specifically,

control panel AI-348 was not designed to prevent water intrusion that resulted in a loss of

power to raw water strainer AC-12B. The NRC performed an initial screening of the finding

in accordance with NRC Manual Chapter IMC 609, Appendix A, The Significance

Determination Process (SDP) for Findings At-Power. Using IMC 0609, Appendix A,

Exhibit 2, Mitigating Systems Screening Questions, dated July 1, 2012, this finding was of

very low safety significance (Green) because it: (1) was not a deficiency affecting the

design or qualification of a mitigating structure, system, or component, and did not result in a

loss of operability or functionality; (2) did not represent a loss of system and/or function;

(3) did not represent an actual loss of function of at least a single train for longer than its

technical specification allowed outage time, or two separate safety systems out-of-service

for longer than their technical specification allowed outage time; (4) did not represent an

actual loss of function of one or more non-technical specification trains of equipment

designated as high safety-significant in accordance with the licensees maintenance rule

program; and (5) did not involve the loss or degradation of equipment or function specifically

designed to mitigate a seismic, flooding or severe weather event. This finding has a cross-

cutting aspect in the area of problem identification and resolution associated with the

organization thoroughly evaluating issues to ensure that resolutions address causes and

extent of conditions commensurate with their safety significance [P.2]. (Section 4OA2.5.f)

Green. A non-cited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control,

was identified involving the failure to accurately model cell level control of river water during

external flooding events. Specifically, the licensee failed to account for losses due to the

E2-5

physical obstructions of trash racks for inflowing river water, the decreased withdrawal rate

of the raw water pumps due to fouling across the traveling screens, and a bounding

inleakage rate for the sluice gates when the river level is at maximum level of 1014 mean

sea level and the intake cell levels are at minimum level of 9769. The licensee entered this

issue into its corrective action program as Condition Report 2014-09155, performed an

operability determination, and initiated action to update station calculations related to intake

cell level control.

This performance deficiency was more than minor, and therefore a finding, because if left

uncorrected, the finding would have the potential to lead to a more significant safety

concern. Specifically, the failure to accurately model flow in and out of the cells could

adversely affect the external flooding mitigation strategy beyond previously identified

equipment capacities and operator actions. This finding was associated with the Mitigating

Systems Cornerstone. The NRC performed an initial screening of the finding in accordance

with NRC Manual Chapter IMC 0609, Appendix A, The Significance Determination

Process (SDP) for Findings At-Power. Using IMC 0609, Appendix A, Exhibit 2, Mitigating

Systems Screening Questions, dated July 1, 2012, this finding is of very low safety

significance (Green) because it: (1) was not a deficiency affecting the design or qualification

of a mitigating system; (2) did not represent a loss of system and/or function; (3) did not

represent an actual loss of function of a single train for greater than its technical

specification allowed outage time; (4) did not represent an actual loss of function of one or

more non-technical specification trains of equipment designated as high safety-significant in

accordance with the licensees maintenance rule program; and (5) did not involve the loss or

degradation of equipment or function specifically designed to mitigate a seismic, flooding or

severe weather event. This finding has a cross-cutting aspect in the area of problem

identification and resolution, operating experience, in that the licensee failed to incorporate

relevant internal operating experience related to previous NRC inspection into

Calculation FC08081 [P.5]. (Section 4OA2.5.g)

Green. A non-cited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control,

was identified involving the failure to translate applicable design requirements into the

specifications for plant systems. Specifically, inadequate design control inputs were used for

analyzing the ability of the vital switchgear room cooling system to perform its safety function

under all conditions. The licensee entered this issue into its corrective action program as

Condition Report 2014-08317 and initiated actions to analyze the ability of vital switchgear

room cooling to meet its specified safety function.

This performance deficiency was more than minor, and therefore a finding, because it

affected the design control attribute of the Mitigating Systems Cornerstone, and it directly

affected the cornerstone objective to ensure availability, reliability, and capability of systems

that respond to initiating events to prevent undesirable consequences. The NRC performed

an initial screening of the finding in accordance with NRC Manual Chapter IMC 0609,

Appendix A, The Significance Determination Process (SDP) for Findings At-Power. Using

IMC 0609, Appendix A, Exhibit 2, Mitigating Systems Screening Questions, dated July 1,

2012, this finding is of very low safety significance (Green) because it: (1) was not a

deficiency affecting the design or qualification of a mitigating system; (2) did not represent a

loss of system and/or function; (3) did not represent an actual loss of function of a single

train for greater than its technical specification allowed outage time; and (4) does not

represent an actual loss of function of one or more non-technical specification trains of

equipment designated as high safety-significant in accordance with the licensees

maintenance rule program for greater than 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . This finding has a cross-cutting aspect

E2-6

in the evaluation component of the problem identification and resolution cross-cutting area

because the licensee failed to thoroughly evaluate issues to ensure that resolutions address

causes and extent of conditions commensurate with their safety significance. Specifically,

the licensee failed to analyze and evaluate a 1998 loss of switchgear cooling event to

ensure that its use as a design assumption bound the worst design basis event [P.2].

(Section 4OA2.5.i)

Green. A cited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, was

identified involving the failure to assure that applicable regulatory requirements and design

bases were correctly translated into specifications, drawings, procedures, and instructions.

Specifically, the licensee failed to properly evaluate NRC Bulletin 88-04, Potential Safety-

Related Pump Loss, for strong pump weak pump interaction regarding auxiliary feedwater

pumps FW-6 and FW-10. The evaluation failed to consider pump-to-pump interaction that

may result due to pump discharge check valve leakage. In addition, the licensee failed to

re-evaluate the condition after surveillance testing performed on November 28, 2010, and

September 1, 2012, identified leakage past both pump discharge check valves. The

licensee entered this issue into its corrective action program as Condition

Report 2014-08381 and initiated actions to re-evaluate NRC Bulletin 88-04.

This performance deficiency was more than minor, and therefore a finding, because it was

associated with the equipment attribute of the Mitigating Systems Cornerstone, and affected

the associated cornerstone objective to ensure the availability, reliability, and capability of

systems that respond to initiating events to prevent undesirable consequences. The NRC

performed an initial screening of the finding in accordance with NRC Manual Chapter

IMC 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-

Power. Using IMC 0609, Appendix A, Exhibit 2, Mitigating Systems Screening Questions,

dated July 1, 2012, the finding was of very low safety significance (Green) because it:

(1) was not a deficiency affecting the design or qualification of a mitigating structure, system,

or component, and did not result in a loss of operability or functionality; (2) did not represent

a loss of system and/or function; (3) did not represent an actual loss of function of at least a

single train for longer than its technical specification allowed outage time, or two separate

safety systems out-of-service for longer than their technical specification allowed outage

time; and (4) did not represent an actual loss of function of one or more non-technical

specification trains of equipment designated as high safety-significant in accordance with

the licensees maintenance rule program. This finding has a cross-cutting aspect in the area

of human performance because the licensee failed to demonstrate a conservative bias in

decision making-practices. Specifically, the licensees determination that the event is not

credible failed to consider documented check valve leakage in the auxiliary feedwater

system [H.14]. (Section 4OA2.5.j)

Green. A cited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, was

identified involving the failure to ensure that the safety-related raw water pumps are

available for safe plant operations down to the design basis low river level. Specifically,

station analysis and abnormal operating procedures would not allow operation of the raw

water pumps to the design basis low river water level. The licensee entered this issue into

its corrective action program as Condition Report 2014-09159 which included actions to re-

evaluate the capability of the raw water pumps to operate at low river levels.

This finding was more than minor, and therefore a finding, because it was associated with

the design control attribute of the Mitigating Systems Cornerstone and affected the

associated cornerstone objective to ensure the availability, reliability, and capability of

E2-7