IR 05000461/2015008

IR 05000461/2015008; 03/02/2015 - 03/20/2015; Clinton Power Station; Evaluations of Changes, Tests, and Experiments and Permanent Plant Modifications

ML15118A496
Person / Time
Site:	Clinton
Issue date:	04/23/2015
From:	Robert Daley Division of Reactor Safety III
To:	Bryan Hanson Exelon Generation Co
References
IR 2015008
Download: ML15118A496 (20)

Text

UNITED STATES ril 23, 2015

SUBJECT:

CLINTON POWER STATION, EVALUATIONS OF CHANGES, TESTS, AND EXPERIMENTS AND PERMANENT PLANT MODIFICATIONS BASELINE INSPECTION REPORT 05000461/2015008

Dear Mr. Hanson:

On March 20, 2015, the U.S. Nuclear Regulatory Commission (NRC) completed an Evaluations of Changes, Tests, and Experiments and Permanent Plant Modifications inspection at your Clinton Power Station. The enclosed inspection report documents the inspection results which were discussed on March 20, 2015, with Mr. Mark Newcomer, 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 subject or severity of the Non-Cited-Violation, you should provide a response within 30 days of the date of this inspection report, with the basis for your denial, to the U.S. Nuclear Regulatory Commission, ATTN: Document Control Desk, Washington, DC 20555-0001, with copies to the Regional Administrator, Region III; the Director, Office of Enforcement, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001; and the NRC Resident Inspector at Clinton 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 Clinton 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 No. 50-461 License No. NPF-62

Enclosure:

Inspection Report 05000461/2015008 w/Attachment: Supplemental Information

REGION III==

Docket No: 50-461 License No: NPF-62 Report No: 05000461/2015008 Licensee: Exelon Generation Company, LLC Facility: Clinton Power Station Location: Clinton, IL Dates: March 2-20, 2015 Inspectors: George M. Hausman, Senior Engineering Inspector (Lead)

James E. Neurauter, Senior Engineering Inspector Lionel Rodriguez, Engineering Inspector Approved by: Robert C. Daley, Chief Engineering Branch 3 Division of Reactor Safety Enclosure

SUMMARY OF FINDINGS

Inspection Report 05000461/2015008; 03/02/2015 - 03/20/2015; Clinton Power Station;

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

This report covers a 2-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: Mitigating Systems

• Severity Level IV-Green. The inspectors identified a finding of very-low safety significance, and an associated Non-Cited Violation of Title 10, Code of Federal Regulations Part 50, Section 59, Changes, Tests and Experiments, (effective January 1, 1997) for a procedure change dated May 2, 1997, where the licensee allowed safety-related switchgear to operate for a limited period of time during plant operation in equipment configurations that were seismically unanalyzed. Specifically, for Safety Evaluation Log 97-060, CPS [Clinton Power Station] Procedure No. 1014.11,

Revision 0, the licensee failed to include a written safety evaluation which provided the bases that concluded for all switchgear configurations that a seismically unanalyzed condition does not involve an unreviewed safety question, and the possibility for a malfunction of a different type than any evaluated previously in the Safety Analysis Report may be created. The licensee entered the issue into their Corrective Action Program as Action Request 02471583, NRC Mod 50.59 Inspection Safety Eval 97-060 for CPS 1014.11, dated March 20, 2015.

The performance deficiency was determined to be more than minor because the finding was associated with the Mitigating Systems cornerstone attribute of protection against external factors and affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). Specifically, switchgear in a seismically unanalyzed condition when relied upon to perform a safety function did not ensure the availability, reliability, or capability of the associated Mitigating Systems to respond to an initiating event such as an earthquake. The inspectors determined that the underlying technical issue was of very-low safety significance (Green) using a detailed risk evaluation. The inspectors did not identify a cross-cutting aspect associated with the finding because the finding was not representative of current performance.

(Section 1R17.1b)

=

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

(71111.17T)

.1 Evaluation of Changes, Tests, and Experiments

a. Inspection Scope

The inspectors reviewed six evaluations performed pursuant to Title 10, Code of Federal Regulations (CFR), Part 50, Section 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 16 screenings, where licensee personnel had determined that a 10 CFR 50.59 evaluation was not necessary. The inspectors reviewed these documents to determine if:

• the changes, tests, 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 16 samples of screenings and/or applicability determinations as defined in Inspection Procedure (IP) 71111.17-04.

b. Findings

Inadequate 50.59 Evaluation for Switchgear in Seismically Unanalyzed Conditions

Introduction:

The inspectors identified a finding of very-low safety significance (Green),

and an associated Severity Level IV, Non-Cited Violation (NCV) of 10 CFR 50.59, Changes, Tests and Experiments, (effective January 1, 1997) for a procedure change dated May 2, 1997, where the licensee allowed safety-related switchgear to operate for a limited period of time during plant operation in equipment configurations that were seismically unanalyzed. Specifically, for Safety Evaluation Log 97-060, CPS

[Clinton Power Station] Procedure No. 1014.11, Revision 0, the licensee failed to include a written safety evaluation which provided the bases that concluded for all switchgear configurations that a seismically unanalyzed condition does not involve an unreviewed safety question, and the possibility for a malfunction of a different type than any evaluated previously in the safety analysis report may be created.

Description:

On February 27, 1997, the licensee generated Condition Report (CR) 1-97-02-273, ABB [ASEA Brown Boveri] and General Electric Breakers Not Seismically Qualified in Racked Out Position. The inspectors noted that the CR and associated Root Cause Report acknowledged that only certain breaker positions had been tested and/or analyzed to seismically qualify the safety-related Division 1, 2, and 3 switchgear.

The CPS Updated Safety Analysis Report (USAR), Section 3.10, Seismic Qualification of Seismic Category I Instrumentation and Electrical Equipment, stated that the requirements of the Institute of Electrical and Electronics Engineers (IEEE) 344, IEEE Recommended Practice for Seismic Qualification of Class 1E Equipment for Nuclear Power Generating Stations and Regulatory Guide (RG) 1.100, Seismic Qualification of Electric and Mechanical Equipment for Nuclear Power Plants, were met for the equipment identified in the USAR Section 3.10. Section 3.10, further stated that per Section 6.1.1, of IEEE 344-75, electrical equipment must be tested on a shake table with mounting and configuration similar to actual service, unless adequate justification can be made to extend the qualification to an untested orientation or configuration.

On March 20, 1997, the licensee completed Risk Evaluation for Seismically Indeterminate Switchgear Configurations, which was included as an attachment to the licensees letter Y-106400 to address the switchgears seismically unanalyzed conditions. The purpose of the evaluation was to address the risk significance of the seismically unanalyzed conditions. The evaluation concluded there were no adverse impacts on the intended safety function of the affected switchgear, and other adjacent cubicles in-service devices (i.e., relays, instruments, etc.); provided the duration of the seismically unanalyzed conditions only existed for a limited period of time. On April 22, 1997, the licensee applied the results of the evaluation and updated the safety analysis report per USAR Change 7-209, Section 3.10, Qualification of Seismic Category I Instrumentation and Electrical Equipment.

On May 2, 1997, the licensee issued Procedure CPS 1014.11, 6900/4160/480V Switchgear/Circuit Breaker Operability Program, which allowed switchgear in a seismically unanalyzed condition to be considered operable for up to 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> as long as administrative controls were implemented. After the 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />, the switchgear was then declared inoperable. The licensees associated Safety Evaluation Log 97-060, CPS Procedure No. 1014.11, Revision 0, concluded that USAR Change 7-209 did not require prior NRC approval.

The inspectors assessed the above changes with respect to the current NEI 96-07, Guidelines for 10 CFR 50.59 Implementation, Revision 1. The guidance in NEI 96-07 was endorsed by the NRC in RG 1.187, Guidance for Implementation of 10 CFR 50.59, Changes, Tests, and Experiments. The inspectors reviewed NEI 96-07, Section 4.3.2, Does the Activity Result in More Than a Minimal Increase in the Likelihood of Occurrence of a Malfunction of an structure, system, or component (SSC) Important to Safety?, which stated that changes in design requirements for earthquakes, tornadoes, and other natural phenomena should be treated as potentially affecting the likelihood of malfunction. The inspectors concluded applying Section 4.3.2, Example 3, that allowing the switchgear to be in a seismically unanalyzed condition required prior NRC approval.

Specifically, when in a seismically unanalyzed configuration, the licensee did not verify the design bases requirement that the switchgear will withstand, without functional impairment, the effects of the safe shutdown earthquake (SSE). Therefore, the seismically unanalyzed configuration resulted in more than a minimal increase in the likelihood of occurrence of a malfunction of a SSC important to safety.

Based on the inspectors review of the licensees Safety Evaluation Log 97-060, the inspectors determined that the licensee incorrectly concluded that switchgear in a seismically unanalyzed condition did not increase the possibility for a malfunction of equipment important to safety evaluated previously in the USAR. The licensees USAR Section 3.10 stated that all Class 1E electrical equipment and instrumentation were designed to withstand, without functional impairment, the effects of the SSE. However, for switchgear in a seismically unanalyzed condition, the licensee did not verify that the equipment would function during and following a postulated SSE event. Therefore, the inspectors concluded a seismically unanalyzed configuration increased the possibility of a switchgear malfunction and was an unreviewed safety question that required prior NRC approval.

The licensee entered the inspectors concern into their Corrective Actions Program (CAP) as Action Request (AR) 02471583, NRC Mod/50.59 Inspection:

Safety Evaluation 97-060 for CPS 1014.11, dated March 20, 2015. The CAP document contained the following recommended corrective actions to address the inspectors concerns:

(1) revise procedure CPS 1014.11 to remove usage of the 48 hour5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> inoperability deferment;

(2) perform a past operability review for exceeding technical specification (TS) action completion times; and

(3) review USAR Section 3.10 for possible changes required to language on breakers in seismically unanalyzed configurations. In addition, AR 02471583 documented creation of Standing Order 2015-02, Actions for Safety Related Breaker Racking Operations, to eliminate entering the 48 hour5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> seismic clock prior to revising Procedure 1014.11.

Analysis:

The inspectors determined that for Safety Evaluation Log 97-060, CPS Procedure No. 1014.11, Revision 0, the licensee failed to include a written safety evaluation which provided the bases that concluded for all switchgear configurations that a seismically unanalyzed condition does not involve an unreviewed safety question, and the possibility for a malfunction of a different type than any evaluated previously in the Safety Analysis Report may be created was contrary to 10 CFR 50.59(d)(1), and was a performance deficiency. Specifically, for switchgear in a seismically unanalyzed condition, the licensee did not verify that the equipment will function during and following a postulated SSE event. Therefore, an unanalyzed configuration created a possibility for a switchgear malfunction of a different type than any previously evaluated during a seismic event, and involved an unreviewed safety question.

The inspectors determined the performance deficiency was more than minor because the finding was associated with the Mitigating Systems cornerstone attribute of protection against external factors, and affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). Specifically, switchgear in a seismically unanalyzed condition when relied upon to perform a safety function did not ensure the availability, reliability, or capability of the associated mitigating systems to respond to an initiating event such as an earthquake.

Violations of 10 CFR 50.59 are dispositioned using the traditional enforcement process instead of the Significance Determination Process (SDP) because they are considered to be violations that potentially impede or impact the regulatory process. This violation is associated with a finding that has been evaluated by the SDP, and communicated with an SDP color reflective of the safety impact of the deficient licensee performance. The SDP, however, does not specifically consider the regulatory process impact. Thus, although related to a common regulatory concern, it is necessary to address the violation and finding using different processes to correctly reflect both the regulatory importance of the violation, and the safety significance of the associated finding.

In this case, the inspectors determined the finding could be evaluated using the SDP in accordance with Inspection Manual Chapter 0609, SDP. Using Attachment 0609.04, Initial Characterization of Findings, Table 2, the inspectors determined that the finding affected the Mitigating Systems cornerstone. As a result, the inspectors evaluated the finding using Appendix A, The SDP for Findings At-Power, Exhibit 2, Mitigating Systems Screening Questions and Exhibit 4, External Events Screening Questions.

The inspectors determined the finding required a detailed risk evaluation because the loss of a switchgear during a seismic event would degrade one or more trains of a system that supports a risk-significant system or function. Specifically, the loss of a Division 1, 2 or 3 switchgear could degrade one train of the emergency power system used to shut the reactor down or maintain it in a safe shutdown condition following a seismic event.

The Senior Reactor Analysts (SRAs) performed a detailed risk evaluation of this issue.

The change in risk for this performance deficiency was assumed to occur following a seismic event with breakers in the divisional switchgear being in an unqualified configuration not allowed by TS. According to the Operations logs, there was one instance in the past 3-years when Division 1 was in a seismically unqualified configuration for approximately 10.8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

The SRAs assumed that a seismically-induced loss of offsite power (LOOP) event would suffice for the initiating event for this issue since the frequencies of seismic LOOP events are based on the lowest fragility SSC (e.g., ceramic insulators). According to information from the NRCs Risk Assessment Standardization Project Tool Box website, the frequency of a seismic LOOP event at Clinton is 5.81E-05/year (based on United States Geological Survey 2008 Hazard Vectors). For the 10.8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> exposure time, this frequency is about 7.2E-08/year. Based on this, the SRAs concluded that the delta-core damage frequency of this performance deficiency is very-low (Green).

In accordance with Section 6.1.d, of the NRC Enforcement Policy this violation is categorized as Severity Level IV, because the resulting changes were evaluated by the SDP as having very-low safety significance (i.e., green finding).

The inspectors did not identify a cross-cutting aspect associated with the finding because the finding was not representative of current performance.

Enforcement:

Title 10 CFR, Part 50, Section 59, Changes, Tests, and Experiments, Subsection (b)(1) (effective January 1, 1997) requires, in part, the licensee to maintain records of changes in procedures to the extent that these changes constitute changes in procedures as described in the Safety Analysis Report. These records must include a written safety evaluation which provides the bases for the determination that the change does not involve an unreviewed safety question.

Title 10 CFR, Part 50, Section 59,Changes. Tests, and Experiments, Subsection (a)(2)

(effective January 1, 1997) states, in part, a proposed change shall be deemed to involve an unreviewed safety question if a possibility for a malfunction of a different type than any evaluated previously in the Safety Analysis Report may be created.

Contrary to the above, from May 2, 1997, to March 20, 2015, for Safety Evaluation Log 97-060, CPS Procedure No. 1014.11, Revision 0, the licensee failed to include a written safety evaluation which provided the bases that concluded for all switchgear configurations that a seismically unanalyzed condition does not involve an unreviewed safety question, and the possibility for a malfunction of a different type than any evaluated previously in the Safety Analysis Report may be created.

This violation is being treated as an NCV, consistent with Section 2.3.2 of the Enforcement Policy because it was a Severity Level IV violation and was entered into the licensees CAP as AR 02471583, NRC Mod 50.59 Inspection Safety Eval 97-060 for CPS 1014.11, dated March 20, 2015. The licensees immediate corrective action created Standing Order Log Number 2015-02, Actions for Safety Related Breaker Racking Operations, to eliminate entering the 48 hour5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> seismic clock until CPS Procedure 1014.11, 6900/4160/480V Switchgear/Circuit Breaker Operability Program, is revised. (NCV 05000461/2015008-01, Inadequate 50.59 Evaluation for Switchgear in Seismically Unanalyzed Conditions)

.2 Permanent Plant Modifications

a. Inspection Scope

The inspectors reviewed nine permanent plant modifications that had been installed in the plant during the last 3 years. 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 nine permanent plant modification samples as defined in IP 71111.17-04.

b. Findings

No findings were identified.

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

No findings were identified.

4OA6 Management Meetings

.1 Exit Meeting Summary

On March 20, 2015, the inspectors presented the inspection results to Mr. Mark Newcomer 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

D. Avery, Regulatory Assurance

P. Bulpitt, Manager, Design Engineering

J. Cunningham, Acting Regulatory Assurance Manager

J. Grim, Engineering

D. Kemper, Operations Director

M. Kimmich, Engineering Support

S. Kowalski, Senior Manager Design Engineering

S. Lakebrink, Sr., Engineering

M. Newcomer, Site Vice-President

J. Peterson, Regulatory Assurance

C. Propst, Work Management Director

D. Shelton, Operations Services Manager

D. Smith, Engineering

J. Smith, Site Engineering Director

T. Stoner, Plant Manager

R. Zacholski, Nuclear Oversight Manager

U.S. Nuclear Regulatory Commission

R. Daley, Chief, Engineering Branch 3, DRS

C. Hunt, Resident Inspector (Acting)

W. Schaup, Senior Resident Inspector

LIST OF ITEMS

Opened and Closed

05000346/2015008-01 NCV Inadequate 50.59 Evaluation for Switchgear in Seismically

Unanalyzed Conditions (Section 1R17.1b.)

Discussed

None