IR 05000373/2013005

IR 05000373-13-005, 05000374-13-005; 10/01/2013 - 12/31/2013; LaSalle County Station, Units 1 and 2; Adverse Weather Protection, Radiological Hazard Assessment and Exposure Controls, Occupational ALARA Planning and Controls, Problem Identif

ML14042A399
Person / Time
Site:	LaSalle
Issue date:	02/11/2014
From:	Michael Kunowski NRC/RGN-III/DRP/B5
To:	Pacilio M Exelon Generation Co, Exelon Nuclear
References
IR-13-005
Download: ML14042A399 (61)

Text

SUBJECT:

LASALLE COUNTY STATION, UNITS 1 AND 2 NRC INTEGRATED INSPECTION REPORT 05000373/2013005; 05000374/2013005

Dear Mr. Pacilio:

On December 31, 2013, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at your LaSalle County Station, Units 1 and 2. The enclosed report documents the inspection results which were discussed on January 8, 2013, with the Site Vice President, Mr. P. Karaba, and other members of your staff.

Six NRC-identified findings of very low safety significance (Green) were identified during this inspection. One of which was also determined to involve a traditional enforcement Severity Level IV violation of NRC requirements.

These findings were determined to involve violations of NRC requirements. The NRC is treating these violations as non-cited violations (NCVs) consistent with Section 2.3.2 of the Enforcement Policy.

If you contest these NCVs, you should provide a response within 30 days of the date of this inspection report, with the basis for your denial, to the Nuclear Regulatory Commission, ATTN:

Document Control Desk, Washington, DC 20555-0001, with copies to the Regional Administrator, Region III; the Director, Office of Enforcement, United States Nuclear Regulatory Commission, Washington, DC 20555-0001; and the NRC Resident Inspector at the LaSalle County Station.

If you disagree with the 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 III, and the NRC Resident Inspector at LaSalle County Station.

February 11, 2014 As a result of the Safety Culture Common Language Initiative, the terminology and coding of cross-cutting aspects were revised beginning in calendar year (CY) 2014. New cross-cutting aspects identified in CY 2014 will be coded under the latest revision to IMC 0310. Cross-cutting aspects identified in the last six months of 2013 using the previous terminology will be converted to the latest revision in accordance with the cross-reference in IMC 0310. The revised cross-cutting aspects will be evaluated for cross-cutting themes and potential substantive cross-cutting issues in accordance with IMC 0305 starting with the CY 2014 mid-cycle assessment review.

In accordance with 10 Code of Federal Regulations 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 Public Document Room or from the Publicly Available Records System (PARS) component of 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/

Michael Kunowski, Chief

Branch 5

Division of Reactor Projects

Docket Nos. 50-373; 50-374 License Nos. NPF-11; NPF-18

Enclosure:

IR 05000373/2013005; 05000374/2013005 w/Attachment: Supplemental Information

REGION III==

Docket Nos:

05000373; 05000374 License Nos:

NPF-11; NPF-18 Report No:

05000373/2013005; 05000374/2013005 Licensee:

Exelon Generation Company, LLC Facility:

LaSalle County Station, Units 1 and 2 Location:

Marseilles, IL Dates:

October 1 through December 31, 2013 Inspectors:

R. Ruiz, Senior Resident Inspector

F. Ramírez, Resident Inspector R. Schultz, IEMA (Illinois Emergency Management Agency) Resident Inspector

T. Go, Health Physicist

J. Cassidy, Senior Health Physicist

B. Palagi, Senior Operations Engineer

R. Baker, Operations Engineer

J. Laughlin, Emergency Preparedness Inspector

Approved by:

M. Kunowski, Chief Branch 5 Division of Reactor Projects

SUMMARY OF FINDINGS

Inspection Report (IR) 05000373/2013005, 05000374/2013005; 10/01/2013 - 12/31/2013;

LaSalle County Station, Units 1 and 2; Adverse Weather Protection, Radiological Hazard Assessment and Exposure Controls, Occupational ALARA Planning and Controls, Problem Identification and Resolution, and Followup of Events and Notices of Enforcement Discretion.

This report covers a 3-month period of inspection by resident inspectors and announced baseline inspections by regional inspectors. Six Green findings were identified by the inspectors. One of which was also determined to involve a traditional enforcement Severity Level-IV violation of NRC requirements. The findings were considered non-cited violations (NCVs) of NRC regulations. The significance of inspection findings is indicated by their color (i.e., greater than Green, or Green, White, Yellow, Red) and determined using Inspection Manual Chapter (IMC) 0609, Significance Determination Process, dated June 2, 2011.

Cross-cutting aspects are determined using IMC 0310, Components Within the Cross Cutting Areas dated October 28, 2011. All violations of NRC requirements are dispositioned in accordance with the NRCs Enforcement Policy dated January 28, 2013. The NRC's program for overseeing the safe operation of commercial nuclear power reactors is described in NUREG-1649, Reactor Oversight Process, Revision

NRC-Identified

and Self-Revealed Findings

Cornerstone: Mitigating Systems

Green.

The inspectors identified a finding of very low safety significance (Green) and associated non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions,

Procedures and Drawings, for the licensees failure to follow the procedure for winter operation preparations. Specifically, all the necessary steps of the procedure were not completed prior to declaring Unit 1 ready for winter operation and, as a result, the bypass blade for the 1B diesel generator (DG) ventilation damper was left in the summer position. Corrective actions planned include revising the Winter Operation Preparation procedure to provide further field guidance on the configuration of ventilation dampers.

The finding was determined to be more than minor because it is associated with the Protection Against External Factors attribute of the Mitigating Systems Cornerstone, and adversely affected the cornerstone objective to ensure availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, with the bypass blade in the summer position, the 1B DG room had the potential to be exposed to extremely cold temperatures without enough margin to compensate with the needed heat, which in turn impacts the availability of the equipment in that room. Using Exhibit 2 of Inspection Manual Chapter (IMC) 0609,

Appendix AProperty "Inspection Manual Chapter" (as page type) with input value "NRC Inspection Manual 0609,</br></br>Appendix A" contains invalid characters or is incomplete and therefore can cause unexpected results during a query or annotation process., The SDP for Findings At-Power, dated June 19, 2012, the finding was determined to have very low safety significance. This finding had a cross-cutting aspect in the area of Human Performance, Work Control, because the licensee did not appropriately coordinate work activities by incorporating actions to address the need to keep personnel apprised of work status. Specifically, the completion and revision of the winter readiness procedure were not sufficiently tracked and, as a result, the licensee missed a procedure step that impacted margin on DG availability (H.3(b)).

(Section 1R01)

Green.

The inspectors identified a finding of very low safety significance (Green) and associated non-cited violation of 10 CFR Part 50, Appendix B, Criterion XI, Test Control, for the licensees failure to ensure that the surveillance testing program was sufficient to demonstrate that safety-related reactor core isolation cooling (RCIC) system components would perform satisfactorily in service. Specifically, the Unit 1 and Unit 2 RCIC steam supply differential pressure switches were not tested at a frequency that assured that they could perform satisfactorily in service, as evidenced by the repeated failures of the components at their quarterly surveillance interval. Corrective actions included the adjustment of the testing periodicity to 46 days, and accelerating the timeline for switch design replacements to address the water intrusion issue.

The inspectors determined the finding could be evaluated using the significance determination process (SDP) in accordance with Inspection Manual Chapter 0609,

Appendix AProperty "Inspection Manual Chapter" (as page type) with input value "NRC Inspection Manual 0609,</br></br>Appendix A" contains invalid characters or is incomplete and therefore can cause unexpected results during a query or annotation process., The Significance Determination Process for Findings At-Power, Exhibit 2, dated June 19, 2012. The inspectors reviewed the Mitigating Systems screening questions in Appendix A, Exhibit 2 and answered No to all the questions, which screened the issue as

Green.

This finding had a cross-cutting aspect in the area of Problem Identification and Resolution, Corrective Action Program, for the failure to take appropriate corrective actions to address adverse trends in a timely manner.

Specifically, in the three-year period from September 2010 to October 2013, these switches failed their surveillances 18 times, primarily due to water intrusion into the switch housing, and corrective actions to increase the frequency or address the water intrusion problem were not taken in a timely manner, commensurate with the safety significance of the issue (P.1(d)). (Section 4OA2.3(2))

Cornerstone: Occupational Radiation Safety

Green.

The inspector identified a finding of very low safety significance (Green) and associated non-cited violation of Technical Specification (TS) 5.7.1.b for the failure to comply with the requirements of the radiation work permit (RWP) during the loading of the multipurpose canister (MPC) containing spent fuel at the refuel floor. The MPC containing the spent fuel bundles was shielded by a HI-TRAC outer shell that also served as the lifting yoke of the loaded MPC. Specifically, the HI-TRAC containing the MPC was serviced by the reactor services personnel and during the preparation of non-destructive dye penetrant testing process, the trunnions part of the lifting yoke device were removed from the HI-TRAC. A worker from the reactor services group on the refuel floor removed the trunnions without the approval of radiation protection (RP)personnel. This constituted a breach of the shielding of the HI-TRAC and MPC. As a result, the worker encountered radiation levels greater than those anticipated, and received unintended dose. The licensees corrective actions included counseling of the involved workers and conducting a stand-down with the reactor services department to reinforce radiological requirements along with communication expectations. The licensee completed an apparent cause evaluation to formulate additional actions to prevent recurrence.

The finding was more than minor because it is associated with the program and process attribute of the Occupational Radiation Safety Cornerstone and adversely affected the cornerstone objective of ensuring adequate protection of worker health and safety from exposure to radiation, in that, access into high radiation areas whose radiological conditions were unknown placed the worker at risk for unnecessary radiation exposure.

The finding was determined to be of very low safety significance because it was not an as-low-as-is-reasonably-achievable (ALARA) planning issue, there was no overexposure or substantial potential for an overexposure, and the licensees ability to assess worker dose was not compromised. The finding involved a cross-cutting aspect in the area of Human Performance, Decision-Making, in that, the worker failed to comply with the RWP requirements for RP hold-points for breaching of a system. Specifically, the decision to remove the trunnions was made without the approval by the RP department and validation that the removal of its parts would not have caused unintended consequences, such as radiation levels greater than anticipated (H.1(b)). Section 2RS1)

Green.

The inspectors identified a finding of very low safety significance (Green) and associated non-cited violation of NRC requirements for the licensees failure to provide the Station ALARA Committee with information needed to ensure that occupational radiation exposure was maintained ALARA. Specifically, TS 5.4.1 requires, in part, that the licensee establish, implement, and maintain applicable procedures recommended in Regulatory Guide 1.33, Revision 2, Appendix A. Section 7 of Appendix A specifies radiation protection procedures for the implementation of ALARA program. Licensee procedures RP-AA-400-1006, Outage Exposure Estimating and Tracking, and RP-AA-401, Operational ALARA Planning and Controls, required effective reviews of the accumulated dose at prescribed intervals to assess the established ALARA controls and to identify/implement prompt dose reduction strategies to minimize exposure to the workers. Contrary to this, the licensee failed to implement steps in these procedures to review the accumulated dose at prescribed intervals on multiple occasions during the L2R14 refueling outage. Corrective actions included instituting appropriate radiological controls and initiating apparent cause evaluations.

The inspectors reviewed Inspection Manual Chapter 0612, Appendix B, Issue Screening and determined that the issue was more than minor because it is associated with the program and process attribute of the Occupational Radiation Safety Cornerstone and adversely affected the associated cornerstone objective, in that, additional radiation exposure was expended during the refueling outage. The inspectors determined that the finding was of very low safety significance in accordance with Inspection Manual Chapter 0609, Appendix C, Occupation Radiation Safety Significance Determination Process. This finding has a cross-cutting aspect in the area of Human Performance, Work Practices, because personnel did not follow procedures (H.4(b)). (Section 2RS2)

Cornerstone: Barrier Integrity

• SL IV/Green. The inspectors identified a finding of very low safety significance (Green)and associated Severity Level IV (SL IV) non-cited violation of 10 CFR 50.71,

Maintenance of records, making of reports. Specifically, 50.71(e) requires that the licensee periodically update the Updated Final Safety Analysis Report (UFSAR), and by letter dated April 12, 2012, the licensee provided the most recent periodic update; however, certain information contained in the update failed to accurately present changes made since the previous submittal and failed to reflect current design and licensing basis. As corrective actions, the licensee has captured the issue in the corrective action program for resolution and plans to review the UFSAR, ensure information is accurately reflected where historical, incorporate the required UFSAR content, and make necessary changes to enhance oversight of future UFSAR revisions.

Consistent with the guidance in Inspection Manual Chapter 0612, Power Reactor Inspection Reports, Appendix B, Issue Screening, dated September 7, 2012, because the issue had the potential to affect the NRC's ability to perform its regulatory function, the violation was reviewed under the traditional enforcement process. Consistent with the guidance in Section 6.9, Paragraph d.9, of the NRC Enforcement Policy, the violation associated with this finding was determined to be a SL-IV violation. Additionally, the significance of the underlying technical issue (i.e., failure to reflect the current licensing basis in the UFSAR, failure to remove obsolete methodology for determining design and operating limits; and, failure to supervise how individuals control and modify information presented in LaSalles UFSAR) was evaluated using the Significance Determination Process (SDP). The finding was determined to be more than minor because, if left uncorrected, it could lead to a more significant safety concern. Specifically, methodologies could be utilized without properly performing 50.59 reviews, leading to revisions to operating ranges, such as those found in the power-to-flow map, which may not have been reflected in the subsequent safety analyses, the failure to update also reduced the ability of the licensee and the NRC to review the stations licensing basis to assess design changes or assess facility safety. The inspectors determined the finding could be evaluated using the SDP in accordance with IMC 0609, Appendix A, The Significance Determination Process for Findings At Power, Exhibit 3, dated June 19, 2012. The inspectors reviewed the Barrier Integrity Screening Questions in Appendix A, Exhibit 3 and answered No to all the questions, which screens the issue as

Green.

The finding had a cross-cutting aspect in the area of Human Performance,

Work Practices, because the licensee failed to supervise and manage the oversight of UFSAR changes with LaSalle station staff, corporate nuclear fuels, and contractors (H.4(c)). (Section 4OA2.3(1))

Green.

The inspectors identified a finding of very low safety significance (Green) and associated non-cited violation of 10 CFR 50, Appendix B, Criterion II, Quality Assurance Program, for the licensees failure to classify secondary containment doors with a quality status of safety-related. Specifically, the licensees failure to classify various secondary containment doors as safety-related was contrary to CC-AA-304,

Component Classification, Revision 5, and was a performance deficiency. The licensees corrective actions were to upgrade the affected doors to the proper safety classification. The licensee entered this issue into its corrective action program as AR 01557738.

The finding was determined to be more than minor because, if left uncorrected, it would become a more significant safety concern. Specifically, by failing to establish and maintain the appropriate quality assurance requirements for these components, the licensee reduced the assurance that initial design, maintenance, and replacement of parts were of sufficient quality to assure reliable service during and following design basis events. The inspectors determined the finding could be evaluated using the Significance Determination Process in accordance with Inspection Manual Chapter 0609, Significance Determination Process, Appendix A, Exhibit 3, Barrier Integrity Screening Questions. The inspectors answered both questions in Section C,

No, therefore, the finding screened as Green because the finding was a qualification deficiency confirmed not to result in loss of operability or functionality. The inspectors did not identify a cross-cutting aspect associated with this finding because the original classification error occurred more than three years ago and is not representative of current performance. (Section 4OA3.3)

REPORT DETAILS

Summary of Plant Status

Unit 1 The unit began the inspection period operating at full power. On November 17, 2013, power was reduced to approximately 75 percent for a control rod sequence exchange. Unit 1 was restored to full power that same day. Additionally, power was reduced to approximately 70 percent on December 14, for a rod sequence exchange, and the unit was returned to full power on December 16.

Unit 2 The unit began the inspection period operating at full power. On October 15, 2013, power was reduced to approximately 80 percent for a control rod sequence exchange. Unit 2 was restored to full power the next day. Additionally, power was reduced to approximately 60 percent on December 7, for a rod sequence exchange, and the unit was returned to full power on December

REACTOR SAFETY

Cornerstone: Initiating Events, Mitigating Systems, Barrier Integrity, and

Emergency Preparedness

1R01 Adverse Weather Protection

.1 Winter Seasonal Readiness Preparations

a. Inspection Scope

The inspectors conducted a review of the licensees preparations for winter conditions to verify that the plants design features and implementation of procedures were sufficient to protect mitigating systems from the effects of adverse weather. Documentation for selected risk-significant systems was reviewed to ensure that these systems would remain functional when challenged by inclement weather. During the inspection, the inspectors focused on plant specific design features and the licensees procedures used to mitigate or respond to adverse weather conditions. Additionally, the inspectors reviewed the UFSAR and performance requirements for systems selected for inspection, and verified that operator actions were appropriate as specified by plant specific procedures. Cold weather protection, such as heat tracing and area heaters, was verified to be in operation where applicable. The inspectors also reviewed corrective action program (CAP) items to verify that the licensee was identifying adverse weather issues at an appropriate threshold and entering them into the CAP in accordance with station corrective action procedures. Documents reviewed are listed in the Attachment to this report. The inspectors reviews focused specifically on the following plant systems due to their risk significance or susceptibility to cold weather issues:

• DG ventilation;

• auxiliary electrical equipment room ventilation; and

• control room ventilation.

This inspection constituted one winter seasonal readiness preparations sample as defined in Inspection Procedure (IP) 71111.01-05.

b. Findings

Failure to Follow Winter Operation Preparations Procedure

Introduction:

The inspectors identified a finding of very low safety significance (Green)and associated non-cited violation (NCV) of 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for the licensees failure to follow the procedure for winter operation preparations. Specifically, all the necessary steps of the procedure were not completed prior to declaring Unit 1 ready for winter operation.

Description:

On November 25, 2013, the inspectors were assessing LaSalles readiness for winter operations. Following the licensees completion of winter readiness activities, the inspectors conducted an independent walkdown to verify that site procedure LOS-AA-A2, Preparation for Winter/Summer Operation, was implemented fully in preparation for the winter months.

During the walkdown, the inspectors noted that, during the licensees winter readiness activities, the bypass blade for a damper in the Unit 1 DG ventilation system had not been placed in the procedurally required state. Specifically, the inspectors found the bypass blade for 1VD01YB in the summer configuration (open) instead of the winter configuration (closed). The purpose of the ventilation damper 1VD01YB was to modulate the air intake to control the temperature in the 1B DG room. If the bypass blade for the damper was in the summer configuration (open), air from the outside could access the 1B DG room even if the damper were closed. If the state of this blade had not been identified by the inspectors, it could have been left open for the entire winter season.

The inspectors reviewed Calculation L-001572, Determination of Required Heating Capacities for the Diesel Generator Buildings. Calculation L-001572 determined the heating load for the DG building and it was bounded by an outside temperature of

-10 degrees Fahrenheit. The calculation concluded that the available heating capacity for the 1B DG room was not sufficient to maintain the room at temperatures above design conditions and recommended that additional heating capacity be added for this purpose. The inspectors concluded that the negative heating margin was further impacted by the fact that the 1VD01YB bypass blade was left in the summer position (open) since the room would have been completely exposed to air coming directly from the outside.

The licensees evaluation of this issue concluded that since multiple equipment operators work on the winter readiness procedure for a period of about 2 months, and there was a procedure revision during that time window, the step to place the 1VD01YB bypass blade in the winter position was missed when passing information from the old revision of the procedure to the new one.

Analysis:

The inspectors determined that the licensees failure to place the bypass blade for ventilation damper 1VD01YB in the winter position as required by site procedure LOS-AA-A2, Preparation for Winter/Summer Operation, was a performance deficiency that warranted further evaluation.

Using the guidance in IMC 0612, Power Reactor Inspection Reports, Appendix B, Issue Screening, dated September 7, 2012, the inspectors determined that the performance deficiency was greater than minor, and therefore a finding, because it was associated with the Protection Against External Factors attribute of the Mitigating Systems Cornerstone, and adversely affected the cornerstone objective to ensure availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, with the bypass blade in the summer position, the 1B DG room had the potential to be exposed to extremely cold temperatures without enough margin to compensate with the needed heat, which in turn negatively impacts the margin to availability of the equipment in that room.

Using Exhibit 2 of IMC 0609, Appendix A, The SDP for Findings At-Power, dated June 19, 2012, the finding was determined to have very low safety significance because all the screening questions were answered No.

This finding had a cross-cutting aspect in the area of Human Performance, Work Control, because the licensee did not appropriately coordinate work activities by incorporating actions to address the need to keep personnel apprised of work status.

Specifically, the completion and revision of the winter readiness procedure were not sufficiently tracked and, as a result, the licensee missed a procedure step that impacted margin on DG availability (H.3(b)).

Enforcement:

Title 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, requires, in part, that activities affecting quality be prescribed by documented instructions, procedures, or drawings, of a type appropriate to the circumstances and be accomplished in accordance with these instructions, procedures, or drawings. Licensee established LOS-AA-A2, Preparation for Winter/Summer Operation, as guidance to ensure that equipment would be protected against extreme winter weather conditions. LOS-AA-A2 requires, in part, that 1VD01YB be placed in its winter configuration between the months of October and April.

Contrary to this requirement, on November 25, 2013, the inspectors identified that the step in procedure LOS-AA-A2 to place the bypass blade for ventilation damper 1VD01YB in the winter position was missed. As a result, the 1B DG room had the potential to be exposed to extremely cold temperatures without enough margin to compensate with the needed heat. Since this issue was entered into the licensees CAP, as AR 1590834, this violation of 10 CFR Part 50, Appendix B, Criterion V, is being treated as an NCV consistent with Section 2.3.2 of the NRC Enforcement Policy (NCV 05000373/2013005-01, Failure to Follow Winter Operation Preparation Procedure).

Corrective actions included closing the affected bypass blade and revising LOS-ZZ-A2 to provide further field guidance on the proper configuration of the damper bypass blades.

1R04 Equipment Alignment

.1 Quarterly Partial System Walkdowns

a. Inspection Scope

The inspector performed a partial system walkdown of the common unit emergency DG cooling water system, which is a risk-significant system. The inspector selected this system based on its risk significance relative to the Reactor Safety Cornerstones at the time it was inspected. The inspectors attempted to identify any discrepancies that could impact the function of the system and, therefore, potentially increase risk. The inspectors reviewed applicable operating procedures, system diagrams, UFSAR, TS requirements, outstanding work orders (WOs), condition reports, and the impact of ongoing work activities on redundant trains of equipment in order to identify conditions that could have rendered the system incapable of performing its intended functions. The inspectors also walked down accessible portions of the system to verify system components and support equipment were aligned correctly and operable.

The inspectors examined the material condition of the components and observed operating parameters of equipment to verify that there were no obvious deficiencies.

The inspectors also verified that the licensee had properly identified and resolved equipment alignment problems that could cause initiating events or impact the capability of mitigating systems or barriers and entered them into the CAP with the appropriate significance characterization. Documents reviewed are listed in the Attachment to this report.

These activities constituted one partial system walkdown sample as defined in IP 71111.04-05.

b. Findings

No findings were identified.

1R05 Fire Protection

.1 Routine Resident Inspector Tours

a. Inspection Scope

The inspectors conducted fire protection walkdowns focused on availability, accessibility, and the condition of firefighting equipment in the following risk-significant plant areas:

• Unit Common Division I DG ventilation equipment room 7366, Fire Zone 7A3;

• Unit 1 reactor building elevation 7106, Fire Zone 2G;

• Unit 2 reactor building elevation 7106, Fire Zone 3G; and

• Unit 2 RCIC/low pressure core spray room, Fire Zone 3I4.

The inspectors reviewed areas to assess if the licensee had implemented a fire protection program that adequately controlled combustibles and ignition sources within the plant, effectively maintained fire detection and suppression capability, maintained passive fire protection features in good material condition, and implemented adequate compensatory measures for out-of-service, degraded, or inoperable fire protection equipment, systems, or features in accordance with the licensees fire plan.

The inspectors selected fire areas based on their overall contribution to internal fire risk as documented in the plants Individual Plant Examination of External Events with later additional insights, their potential to impact equipment which could initiate or mitigate a plant transient, or their impact on the plants ability to respond to a security event.

Using the documents listed in the Attachment to this report, the inspectors verified that fire hoses and extinguishers were in their designated locations and available for immediate use; that fire detectors and sprinklers were unobstructed; that transient material loading was within the analyzed limits; and fire doors, dampers, and penetration seals appeared to be in satisfactory condition. The inspectors also verified that minor issues identified during the inspection were entered into the licensees CAP.

Documents reviewed are listed in the Attachment to this report.

These activities constituted four quarterly fire protection inspection samples as defined in IP 71111.05-05.

b. Findings

No findings were identified.

1R06 Flooding

.1 Internal Flooding

a. Inspection Scope

The inspectors reviewed selected risk important plant design features and licensee procedures intended to protect the plant and its safety-related equipment from internal flooding events. The inspectors reviewed flood analyses and design documents, including the UFSAR, engineering calculations, and abnormal operating procedures, to identify licensee commitments. In addition, the inspectors reviewed licensee drawings to identify areas and equipment that may be affected by internal flooding caused by the failure or misalignment of nearby sources of water, such as the fire suppression or the circulating water systems. The inspectors also reviewed the licensees CAP documents with respect to past flood-related items identified in the CAP to verify the adequacy of the corrective actions. The inspectors performed a walkdown of the following plant area to assess the adequacy of watertight doors and verify drains and sumps were clear of debris and were operable, and that the licensee complied with its commitments:

• evaluated LOA-FLD-001 procedure and walked down reactor building to evaluate station physical readiness for procedure execution.

Documents reviewed are listed in the Attachment to this report.

This inspection constituted one internal flooding sample as defined in IP 71111.06-05.

b. Findings

No findings were identified.

1R07 Annual Heat Sink Performance

.1 Heat Sink Performance

a. Inspection Scope

The inspectors reviewed the licensees testing of DG cooling water heat exchangers to verify that potential deficiencies did not mask the licensees ability to detect degraded performance, to identify any common cause issues that had the potential to increase risk, and to ensure that the licensee was adequately addressing problems that could result in initiating events that would cause an increase in risk. The inspectors reviewed the licensees observations as compared against acceptance criteria, the correlation of scheduled testing and the frequency of testing, and the impact of instrument inaccuracies on test results. Inspectors also verified that test acceptance criteria considered differences between test conditions, design conditions, and testing conditions. Documents reviewed are listed in the Attachment to this report.

This annual heat sink performance inspection constituted one sample as defined in IP 71111.07-05.

b. Findings

No findings were identified.

1R11 Licensed Operator Requalification Program

.1 Resident Inspector Quarterly Review of Licensed Operator Requalification

a. Inspection Scope

On October 25, 2013, the inspectors observed a crew of licensed operators in the plants simulator during licensed operator requalification training to verify that operator performance was adequate, evaluators were identifying and documenting crew performance problems, and training was being conducted in accordance with licensee procedures. The inspectors evaluated the following areas:

• licensed operator performance;

• crews clarity and formality of communications;

• ability to take timely actions in the conservative direction;

• prioritization, interpretation, and verification of annunciator alarms;

• correct use and implementation of abnormal and emergency procedures;

• control board manipulations;

• oversight and direction from supervisors; and

• ability to identify and implement appropriate TS actions and Emergency Plan actions and notifications.

The crews performance in these areas was compared to pre-established operator action expectations and successful critical task completion requirements. Documents reviewed are listed in the Attachment to this report.

This inspection constituted one quarterly licensed operator requalification program simulator sample as defined in IP 71111.11.

b. Findings

No findings were identified.

.2 Resident Inspector Quarterly Observation of Heightened Activity or Risk

a. Inspection Scope

On October 18, 2013, the inspectors observed Unit 2 turbine control valve and bypass valve testing in the control room as well as observed the associated return to full power.

This was an activity that required heightened awareness or was related to increased risk. The inspectors evaluated the following areas:

• licensed operator performance;

• crews clarity and formality of communications;

• ability to take timely actions in the conservative direction;

• prioritization, interpretation, and verification of annunciator alarms (if applicable);

• correct use and implementation of procedures;

• control board (or equipment) manipulations;

• oversight and direction from supervisors; and

• ability to identify and implement appropriate TS actions and Emergency Plan actions and notifications (if applicable).

The performance in these areas was compared to pre-established operator action expectations, procedural compliance and task completion requirements. Documents reviewed are listed in the Attachment to this report.

This inspection constituted one quarterly licensed operator heightened activity/risk sample as defined in IP 71111.11.

b. Findings

No findings were identified.

.3 Annual Testing Results

Biennial Written and Annual Operating Test Results (71111.11A)

a. Inspection Scope

The inspectors reviewed the overall pass/fail results of the Biennial Written Examination, and the Annual Operating Test, administered by the licensee from October 14 through November 22, 2013, and required by 10 CFR 55.59(a). The results were compared to the thresholds established in Inspection Manual Chapter 0609, Appendix I, Licensed Operator Requalification Significance Determination Process (SDP)," dated December 6, 2011, to assess the overall adequacy of the licensees Licensed Operator Requalification Training (LORT) Program to meet the requirements of 10 CFR 55.59.

(02.02)

This inspection constituted one annual licensed operator requalification examination results sample as defined in IP 71111.11-05.

b. Findings

No findings were identified.

.4 Biennial Review

a. Inspection Scope

The following inspection activities were conducted during the weeks of November 11 and 18, 2013, to assess: 1) the effectiveness and adequacy of the facility licensees implementation and maintenance of its systems approach to training (SAT) based LORT Program, put into effect to satisfy the requirements of 10 CFR 55.59; 2) conformance with the requirements of 10 CFR 55.46 for use of a plant referenced simulator to conduct operator licensing examinations and for satisfying experience requirements; and 3) conformance with the operator license conditions specified in 10 CFR 55.53. The documents reviewed are listed in the Attachment to this report.

• Licensee Requalification Examinations (10 CFR 55.59(c); SAT Element 4 as Defined in 10 CFR 55.4): The inspectors reviewed the licensees program for development and administration of the LORT biennial written examination and annual operating tests to assess the licensees ability to develop and administer examinations that were acceptable for meeting the requirements of 10 CFR 55.59(a).

- The inspectors conducted a detailed review of two biennial requalification written examination versions to assess content, level of difficulty, and quality of the written examination materials. (02.03)

- The inspectors conducted a detailed review of ten job performance measures (JPMs) and six dynamic simulator scenarios to assess content, level of difficulty, and quality of the operating test materials. (02.04)

- The inspectors observed the administration of the annual operating test to assess the licensees effectiveness in conducting the examination(s),including the conduct of pre-examination briefings, evaluations of individual operator and crew performance, and post-examination analysis. The inspectors evaluated the performance of two simulator crews in parallel with the facility evaluators during four dynamic simulator scenarios and evaluated various licensed crew members concurrently with facility evaluators during the administration of several JPMs. (02.05)

- The inspectors assessed the adequacy and effectiveness of the remedial training conducted since the last requalification examinations and the training planned for the current examination cycle to ensure that they addressed weaknesses in licensed operator or crew performance identified during training and plant operations. The inspectors reviewed remedial training procedures and individual remedial training plans. (02.07)

• Conformance with Examination Security Requirements (10 CFR 55.49): The inspectors conducted an assessment of the licensees processes related to examination physical security and integrity (e.g., predictability and bias) to verify compliance with 10 CFR 55.49, Integrity of Examinations and Tests. The inspectors reviewed the facility licensees examination security procedure, and observed the implementation of physical security controls (e.g., access restrictions and simulator I/O controls) and integrity measures (e.g., security agreements, sampling criteria, bank use, and test item repetition) throughout the inspection period. (02.06)

• Conformance with Operator License Conditions (10 CFR 55.53): The inspectors reviewed the facility licensee's program for maintaining active operator licenses and to assess compliance with 10 CFR 55.53(e) and (f). The inspectors reviewed the procedural guidance and the process for tracking on-shift hours for licensed operators, and which control room positions were granted watch-standing credit for maintaining active operator licenses. (02.08)

• Conformance with Simulator Requirements Specified in 10 CFR 55.46: The inspectors assessed the adequacy of the licensees simulation facility (simulator)for use in operator licensing examinations and for satisfying experience requirements. The inspectors reviewed a sample of simulator performance test records (e.g., transient tests, malfunction tests, scenario-based tests, post-event tests, steady-state tests, and core performance tests), simulator discrepancies, and the process for ensuring continued assurance of simulator fidelity in accordance with 10 CFR 55.46. The inspectors reviewed and evaluated the discrepancy corrective action process to ensure that simulator fidelity was being maintained. Open simulator discrepancies were reviewed for importance relative to the impact on 10 CFR 55.45 and 55.59 operator actions, as well as on nuclear and thermal hydraulic operating characteristics. (02.09)

• Problem Identification and Resolution (10 CFR 55.59(c): SAT Element 5 as Defined in 10 CFR 55.4): The inspectors assessed the licensees ability to identify, evaluate, and resolve problems associated with licensed operator performance (a measure of the effectiveness of its LORT Program and the ability to implement appropriate corrective actions to maintain the LORT Program up-to-date). The inspectors reviewed documents related to licensed operator performance issues (e.g., recent examination and inspection reports including cited and Non-Cited Violations; NRC End-of-Cycle and Mid-Cycle reports; NRC plant issue matrix; licensee event reports; licensee condition/problem identification reports, including documentation of plant events and review of industry operating experience). The inspectors also sampled the licensees quality assurance oversight activities, including licensee training department self-assessment reports. (02.10)

This inspection constituted one Biennial Licensed Operator Requalification Program inspection sample as defined in IP 71111.11-05.

b. Findings

No findings were identified.

1R12 Maintenance Effectiveness

.1 Routine Quarterly Evaluations

a. Inspection Scope

The inspectors evaluated degraded performance issues involving the following risk-significant systems:

• nuclear fuel; and

• B.5.b equipment.

The inspectors reviewed events, such as where ineffective equipment maintenance had resulted in valid or invalid automatic actuations of engineered safeguards systems, and independently verified the licensee's actions to address system performance or condition problems in terms of the following:

• implementing appropriate work practices;

• identifying and addressing common cause failures;

• scoping of systems in accordance with 10 CFR 50.65(b) of the maintenance rule;

• characterizing system reliability issues for performance;

• charging unavailability for performance;

• trending key parameters for condition monitoring;

• ensuring 10 CFR 50.65(a)(1) or (a)(2) classification or re-classification; and

• verifying appropriate performance criteria for structures, systems, and components (SSCs)/functions classified as (a)(2), or appropriate and adequate goals and corrective actions for systems classified as (a)(1).

The inspectors assessed performance issues with respect to the reliability, availability, and condition monitoring of the system. In addition, the inspectors verified maintenance effectiveness issues were entered into the CAP with the appropriate significance characterization. Documents reviewed are listed in the Attachment to this report.

This inspection constituted two quarterly maintenance effectiveness samples as defined in IP 71111.12-05.

b. Findings

No findings were identified.

1R13 Maintenance Risk Assessments and Emergent Work Control

.1 Maintenance Risk Assessments and Emergent Work Control

a. Inspection Scope

The inspectors reviewed the licensee's evaluation and management of plant risk for the maintenance and emergent work activity affecting risk-significant and safety-related equipment listed below to verify that the appropriate risk assessments were performed prior to removing equipment for work:

• effects of tornado warning on risk status and work schedule.

This activity was selected based on its potential risk significance relative to the Reactor Safety Cornerstones. As applicable for the activity, the inspectors verified that risk assessments were performed as required by 10 CFR 50.65(a)(4) and were accurate and complete. When emergent work was performed, the inspectors verified that the plant risk was promptly reassessed and managed. The inspectors reviewed the scope of maintenance work, discussed the results of the assessment with the licensee's probabilistic risk analyst or shift technical advisor, and verified plant conditions were consistent with the risk assessment. The inspectors also reviewed TS requirements and walked down portions of redundant safety systems, when applicable, to verify risk analysis assumptions were valid and applicable requirements were met.

Documents reviewed are listed in the Attachment to this report. This maintenance risk assessment and emergent work control activity constituted one sample as defined in IP 71111.13-05.

b. Findings

No findings were identified.

1R15 Operability Determinations and Functional Assessments

.1 Operability Evaluations

a. Inspection Scope

The inspectors reviewed the following issues:

• Unit 1 V automatic depressurization system safety relief valve (SRV);

• Unit 2 A DG cooling water pump pressure and flow degradation;

• Unit 2 C SRV leakby evaluation; and

• drywell leakage instrumentation degradation.

The inspectors selected these potential operability issues based on the risk significance of the associated components and systems. The inspectors evaluated the technical adequacy of the evaluations to ensure that TS operability was properly justified and the subject component or system remained available such that no unrecognized increase in risk occurred. The inspectors compared the operability and design criteria in the appropriate sections of the TSs and UFSAR to the licensees evaluations to determine whether the components or systems were operable. Where compensatory measures were required to maintain operability, the inspectors determined whether the measures in place would function as intended and were properly controlled. The inspectors determined, where appropriate, compliance with bounding limitations associated with the evaluations. Additionally, the inspectors reviewed a sampling of CAP documents to verify that the licensee was identifying and correcting any deficiencies associated with operability evaluations. Documents reviewed are listed in the Attachment to this report.

This operability inspection constituted four samples as defined in IP 71111.15-05.

b. Findings

No findings were identified.

1R18 Plant Modifications

.1 Plant Modifications

a. Inspection Scope

The inspectors reviewed the licensees change to the drywell and reactor building oxygen monitoring procedure.

The inspectors reviewed the changes and associated 10 CFR 50.59 safety evaluation screening against the design basis, the UFSAR, and the TS, as applicable, to verify that the modification did not affect the operability or availability of the affected systems. The inspectors, as applicable, observed ongoing and completed work activities to ensure that the modifications were installed as directed and consistent with the design control documents; the modifications operated as expected; post-modification testing adequately demonstrated continued system operability, availability, and reliability; and that operation of the modifications did not impact the operability of any interfacing systems. As applicable, the inspectors verified that relevant procedure, design, and licensing documents were properly updated. Lastly, where applicable, the inspectors discussed the plant modification with operations, engineering, and training personnel to ensure that the individuals were aware of how the operation with the plant modification in place could impact overall plant performance. Documents reviewed are listed in the to this report.

This inspection constituted one permanent plant modification sample as defined in IP 71111.18-05.

b. Findings

No findings were identified.

1R19 Post-Maintenance Testing

.1 Post-Maintenance Testing

a. Inspection Scope

The inspectors reviewed the following post-maintenance activities to verify that procedures and test activities were adequate to ensure system operability and functional capability:

• Unit 1 B DG work window;

• Unit 2 Division I residual heat removal (RHR) service water work window;

• Unit 1 chemical cleaning of 1VY03A, core standby cooling system equipment ventilation cooler;

• Unit Common B train of the auxiliary electrical equipment room ventilation system, following emergent repair of fan motor; and

• Unit 1 Division III outage.

These activities were selected based upon the structure, system, or component's ability to impact risk. The inspectors evaluated these activities for the following (as applicable):

the effect of testing on the plant had been adequately addressed; testing was adequate for the maintenance performed; acceptance criteria were clear and demonstrated operational readiness; test instrumentation was appropriate; tests were performed as written in accordance with properly reviewed and approved procedures; equipment was returned to its operational status following testing (temporary modifications or jumpers required for test performance were properly removed after test completion); and test documentation was properly evaluated. The inspectors evaluated the activities against TSs, the UFSAR, 10 CFR Part 50 requirements, licensee procedures, and various NRC generic communications to ensure that the test results adequately ensured that the equipment met the licensing basis and design requirements. In addition, the inspectors reviewed CAP documents associated with post-maintenance tests to determine whether the licensee was identifying problems and entering them in the CAP and that the problems were being corrected commensurate with their importance to safety.

Documents reviewed are listed in the Attachment to this report.

This inspection constituted five post-maintenance testing samples as defined in IP 71111.19-05.

b. Findings

No findings were identified.

1R22 Surveillance Testing

.1 Surveillance Testing

a. Inspection Scope

The inspectors reviewed the test results for the following activities to determine whether risk-significant systems and equipment were capable of performing their intended safety function and to verify testing was conducted in accordance with applicable procedural and TS requirements:

• Unit 1 A DG fast start (Routine); and

• Unit 2 reactor high dome pressure relay response time testing (Routine).

The inspectors observed in-plant activities and reviewed procedures and associated records to determine the following:

• did preconditioning occur;

• the effects of the testing were adequately addressed by control room personnel or engineers prior to the commencement of the testing;

• acceptance criteria were clearly stated, demonstrated operational readiness, and were consistent with the system design basis;

• plant equipment calibration was correct, accurate, and properly documented;

• as-left setpoints were within required ranges; and the calibration frequency was in accordance with TSs, the UFSAR, procedures, and applicable commitments;

• measuring and test equipment calibration was current;

• test equipment was used within the required range and accuracy; applicable prerequisites described in the test procedures were satisfied;

• test frequencies met TS requirements to demonstrate operability and reliability; tests were performed in accordance with the test procedures and other applicable procedures; jumpers and lifted leads were controlled and restored where used;

• test data and results were accurate, complete, within limits, and valid;

• test equipment was removed after testing;

• where applicable for inservice testing activities, testing was performed in accordance with the applicable version of Section XI, American Society of Mechanical Engineers code, and reference values were consistent with the system design basis;

• where applicable, test results not meeting acceptance criteria were addressed with an adequate operability evaluation or the system or component was declared inoperable;

• where applicable for safety-related instrument control surveillance tests, reference setting data were accurately incorporated in the test procedure;

• where applicable, actual conditions encountering high resistance electrical contacts were such that the intended safety function could still be accomplished;

• prior procedure changes had not provided an opportunity to identify problems encountered during the performance of the surveillance or calibration test;

• equipment was returned to a position or status required to support the performance of its safety functions; and

• all problems identified during the testing were appropriately documented and dispositioned in the CAP.

Documents reviewed are listed in the Attachment to this report.

This inspection constituted two routine surveillance testing samples as defined in IP 71111.22-02.

b. Findings

No findings were identified.

1EP4 Emergency Action Level and Emergency Plan Changes

a. Inspection Scope

The Office of Nuclear Security and Incident Response headquarters staff performed an in-office review of the latest revisions to the Emergency Plan and various Emergency Plan Implementing Procedures. The licensee transmitted the Emergency Plan Implementing Procedures revisions to the NRC pursuant to the requirements of 10 CFR Part 50, Appendix E, Section V, Implementing Procedures. The NRC review was not documented in a safety evaluation report and did not constitute approval of licensee-generated changes; therefore, this revision is subject to future inspection. The documents reviewed during this inspection are listed in the Attachment to this report.

This emergency action level and emergency plan change inspection constituted one sample as defined in IP 71114.04-05.

b. Findings

No findings were identified.

2RS1 Radiological Hazard Assessment and Exposure Controls

The inspection activities supplement those documented in NRC Inspection Reports 05000373(374)/2013002 and 05000373(374)/2013003, and constituted one complete sample as defined in IP 71124.01-05.

.1 Radiological Hazard Assessment (02.02)

a. Inspection Scope

The inspectors observed work in potential airborne areas and evaluated whether the air samples were representative of the breathing air zone. The inspectors evaluated whether continuous air monitors were located in areas with low background to minimize false alarms and were representative of actual work areas. The inspectors evaluated the licensees program for monitoring levels of loose surface contamination in areas of the plant with the potential for the contamination to become airborne.

b. Findings

No findings were identified.

.2 Radiological Hazards Control and Work Coverage (02.05)

a. Inspection Scope

The inspectors evaluated ambient radiological conditions (e.g., radiation levels or potential radiation levels) during tours of the facility. The inspectors assessed whether the conditions were consistent with applicable posted surveys, RWPs, and worker briefings.

The inspectors assessed whether radiation monitoring devices were placed on the individuals body consistent with licensee procedures. The inspectors assessed whether the dosimeter was placed in the location of highest expected dose or that the licensee properly employed an NRC-approved method of determining effective dose equivalent.

The inspectors examined the posting and physical controls for selected high radiation areas and very high radiation areas to verify conformance with the occupational performance indicator.

b. Findings

No findings were identified.

.3 Radiation Worker Performance (02.07)

a. Inspection Scope

The inspectors observed radiation worker performance with respect to stated RP work requirements. The inspectors assessed whether workers were aware of the radiological conditions in their workplace and the RWP controls/limits in place, and whether their performance reflected the level of radiological hazards present.

b. Findings

Failure to Comply with Radiation Work Permit Requirements

Introduction:

The inspector identified a finding of very low safety significance (Green)and an associated non-cited Violation of TS 5.7.1.b for the licensees failure to comply with the requirements of the RWP associated with the loading of the multi-purpose canister (MPC) containing spent fuel at the refuel floor. As a result, a worker encountered radiation levels greater than those anticipated.

Description:

The inspectors reviewed an incident revealed to the licensee through an alarming electronic dosimeter that occurred on May 22, 2013, during a completion of dry cask storage activities during the removal of the trunnions system of the HI-TRAC/MPC loaded with spent fuel bundles. The inspectors reviewed the licensees condition report, associated radiation survey data and RWP, and follow-up human performance evaluation and discussed the incident with members of the reactor services group, RP staff, and the licensees management.

On May 22, 2013, the reactor services personnel at LaSalle removed the trunnions in preparation for non-destructive testing. However, the trunnions, parts of lifting yoke, were removed from the HI-TRAC without the approval of the RP staff. The removal represented a breaching or opening of process line and ultimately reduced the amount of radiation shielding from the spent fuel inside the HI-TRAC/MPC.

With the MPC loaded with fuel and sitting on the decontamination pad, the trunnions were removed to perform a non-destructive dye penetrant test. The trunnions are threaded into the side of the HI-TRAC and are used to lift the HI-TRAC with the lifting yoke. Although not their design function, the trunnions also provide additional shielding from the spent fuel when installed in the HI-TRAC. Once the trunnions were removed, the amount of shielding provided was reduced and one particular worker of the reactor services group was exposed to a higher dose rate than expected. A follow-up survey indicated a dose rate of 600 milliRem/hour at the opening, and 70 milliRem/hour at 30 centimeters from the opening. This resulted in the worker receiving a total dose of 9 milliRem during the activity. This was an unplanned activity that deviated from the RWP. Routinely, the HI-TRAC was stored in the dry cask storage building when no dry cask storage campaign was in-progress. The non-destructive dye penetrant test on the trunnions was scheduled to be completed while the HI-TRAC was empty and located in the dry cask storage building. However, a decision was made by reactor services personnel, independent of the RP department, to transport the HI-TRAC onto the refuel floor and to remove the trunnions on the refuel floor.

Radiation Work Permit No. 100014638, Dry Cask Storage Activities, governed dry cask storage work in 2013 and included special instructions to workers. The instructions included the requirements in RP-AA-401, Exposure Reduction Measures, page 3 of 17, Exposure Reduction Measures of Item Nos. 11 and 12, that state, It is mandatory that RP staff will be informed of all activities associated with this project and RP staff be informed of any deviation from the original scope of work to be performed. However, the licensee RP staff was neither informed of all activities associated with the change of plan by the reactor services department nor did reactor services personnel perform the necessary notifications that there was a deviation from the original scope of work associated with the cask loading and removal of the trunnions.

Reactor services personnel made the decision to remove the trunnions during the morning pre-job briefing. Radiation protection personnel were not present at this briefing. Consequently, the change in work plan to remove the trunnions was made without further planning and support from the RP department. This reactor services plan deviated from the original scope, and according to the licensees investigation, the reactor services staff lack of knowledge of the shielding and radiation characteristics of the loaded HI-TRAC contributed to the event. The licensee assumed that there were no issues of radiological significance and applied the same level of rigor as other routine dry cask campaign in terms of radiological hold-points. Specifically, the reactor services personnel thought that the trunnions were easily handled and accessible when the HI-TRAC was sitting on the pad at the refuel floor with the scaffolding in place. During the removal of the second trunnions; however, the worker received a dose rate alarm.

Dose rate surveys were performed by an RP technician immediately after the event and the trunnions were reinstalled on the HI-TRAC. All proper notifications were made and an investigation was started on this event.

Analysis:

The inspectors determined that the failure to properly evaluate the radiological hazards associated with the unplanned removal of the trunnions from the HI-TRAC on the refuel floor was a performance deficiency in accordance with IMC 0612, "Power Reactor Inspection Reports," Appendix B, "Issue Screening," dated September 7, 2012.

Additionally, the inspectors determined that the performance deficiency was of more than minor significance. Specifically, if left uncorrected, the performance deficiency had the potential to lead to a more significant safety concern, in that, not evaluating the radiological impact and controlling personnel exposures associated with the unplanned removal of the trunnions resulted in unnecessary and unplanned elevation of ambient radiation fields where the worker was present. Additionally, there was industry operating experiences on controlling opening of process line or breaching an unknown system.

Consequently, the inspectors also concluded that this activity was within the licensees ability to foresee and should have been prevented. The finding was not subject to traditional enforcement since the incident did not impact the NRCs ability to perform its regulatory function and was not willful.

Since the finding involved occupational radiation safety, the inspectors utilized IMC 0609, Appendix C, Occupational Radiation Safety Significance Determination Process, dated August 19, 2008, to assess its significance. The inspectors determined that the finding did not involve an overexposure, a substantial potential for an over exposure, or a compromised ability to assess dose. Also, the finding did not involve ALARA planning or work controls issue that resulted in excessive collective radiation exposures. Consequently, the inspectors determined that the finding was of very low safety significance (Green).

The finding involved a cross-cutting aspect in the area of Human Performance, Decision-Making, in that, licensee personnel did not demonstrate RWP compliance to RP hold-points before breaching a system, i.e., removing a component that provided radiation shielding. Specifically, the decision to remove the trunnions from the HI-TRAC was made without the approval by the RP department and to validation that the removal of its parts would not have caused unintended consequences, such as radiation levels greater than anticipated (H.1(b)). The licensees reactor services personnel assumed that there was no consequence to radiological conditions associated with the removal of the trunnions from the HI-TRAC/MPC.

Enforcement:

Technical Specification 5.7.1.b requires that access to and activities in, high radiation areas be controlled by means of an RWP or equivalent that includes specification of radiation dose rates in the immediate work area(s) and other appropriate RP equipment and measures. Radiation Work Permit No. 100014638, Dry Cask Storage Activities, governed dry cask storage work in 2013 that included special instructions from procedure RP-AA-401, Exposure Reduction Measures, that RP staff be informed of all activities associated with the project and of any deviations from the original scope of work to be performed.

Contrary to the RWP, on May 22, 2013, a worker removing the trunnions from a HI-TRAC/MPC that was loaded with spent fuel bundles was exposed to unexpected radiation levels. The worker failed to inform the RP staff of work activities including the details of the project and procedural deviations from the original scope of work to be performed.

Since the failure to comply with TSs was of very low safety significance, corrective actions were taken as described above, and the issue was entered into the licensees CAP (as AR 01516861), the violation is being treated as an NCV consistent with Section 2.3.2 of the NRC Enforcement Policy (NCV 05000373/2013005-02; 05000374/2013005-02, Failure to Comply with Radiation Work Permit Requirements).

2RS2 Occupational As-Low-As-Is-Reasonably-Achievable Planning and Controls

The inspection activities supplement those documented in NRC Inspection Reports 05000373(374)/2012002 and 05000373(374)/2013002, and constituted one complete sample as defined in IP 71124.02-05.

.1 Radiological Work Planning (02.02)

a. Inspection Scope

The inspectors compared the results achieved (dose rate reductions, person-rem used)with the intended dose established in the licensees ALARA planning for select work activities. The inspectors compared the person-hour estimates provided by maintenance planning and other groups to the RP group with the actual work activity time requirements, and evaluated the accuracy of these time estimates. The inspectors assessed the reasons (e.g., failure to adequately plan the activity, failure to provide sufficient work controls) for any inconsistencies between intended and actual work activity doses.

The inspectors determined whether post-job reviews were conducted and if identified problems were entered into the licensees CAP.

b. Findings

Failure to Provide Station ALARA Committee (SAC) with Necessary Information to Ensure Occupational Radiation Exposure Is Maintained ALARA

Introduction:

The inspectors identified a finding of very low safety significance (Green)and associated NCV of NRC requirements for the licensees failure to provide the SAC with information needed to ensure that occupational radiation exposure was maintained ALARA.

Description:

The inspectors identified events and conditions that were present during the previous refueling outage, L2R14, that were not presented to the SAC in a timely manner to develop and implement dose mitigation strategies.

An ALARA program integrates management philosophy and regulatory requirements.

Some objectives for maintaining occupational exposure ALARA are:

• Maintain individual and collective doses to workers and the general public ALARA

• Develop a comprehensive exposure reduction program, which provides a means for identifying and implementing new and proven methods for minimizing exposure.

• Incorporate management commitment to dose minimization into daily plant operations, both outage and non-outage work activities.

The refueling outage started on February 11, 2013, with a dose goal of 153 rem for all scheduled work activities. The outage ended just ahead of schedule on March 4, 2013, but the licensee expended over 284 rem to complete the work activities. The inspectors assessed the overall ALARA performance at the time of the performance deficiency and found the three-year rolling average was below 240 rem (2009-2011).

The performance deficiency occurred because of a failure to follow station procedures of RP-AA-400-1006, Outage Exposure Estimating and Tracking and RP-AA-401, Operational ALARA Planning and Controls, specifically:

• Step 4.5.3 of RP-AA-400-1006 states that exposure estimates should be re-evaluated and adjusted if needed and this review should include the SAC.

This review is important because the SAC needs to be informed of activities that are at risk of not meeting dose estimates (a regulatory risk) and it is an opportunity to gain SAC approval of dose reduction strategies or approval of abort/contingency actions. This review was not completed in the 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />; rather, it was 10 days or midway through the outage, well after increased dose rates were evident in the plant.

• Step 4.2.4 of RP-AA-401 requires Work In Progress (WIP) reviews to be completed at 25 percent, 50 percent, and 80 percent of progress. The inspectors noted that some WIPs were not initiated until the RWP exceeded 60 percent or 80 percent of the allotted dose before reviews identified increased dose rates in the work areas. Another WIP exceeded 100 percent of the estimated dose before the increased dose rates were accounted for and/or it was recognized that additional work was performed on the RWP. Effective evaluation of work progress is important to assess the effectiveness of established ALARA controls and to identify/implement prompt dose reduction strategies to minimize exposure to the workers.

Analysis:

The inspectors determined that the failure to provide the SAC with information needed to ensure that occupational radiation exposure was maintained ALARA was a performance deficiency of more than minor significance, and thus a finding, in accordance with IMC 0612, "Power Reactor Inspection Reports," Appendix B, "Issue Screening," dated September 7, 2012. Specifically, the performance deficiency was associated with the Program and Process attribute of the Occupational Radiation Safety Cornerstone and the performance deficiency adversely affected the associated cornerstone objective, in that, additional allotted dose from radiation exposure was expended during the refueling outage. The inspectors also concluded that this activity was within the licensees ability to foresee and should have been prevented. Since the finding involved occupational radiation safety, the inspectors utilized IMC 0609, Appendix C, Occupational Radiation Safety Significance Determination Process, dated August 19, 2008, to assess its significance. The inspectors determined that the finding did not involve an overexposure; a substantial potential for an overexposure; a compromised ability to assess dose; or unplanned, unintended occupational collective dose. Consequently, the inspectors determined that the finding was of very low safety significance (Green).

As described above, the finding was caused by the failure to follow established procedures. Consequently, the inspectors determined that the finding has a cross-cutting aspect in the area of Human Performance, Work Practices, because personnel did not follow procedures (H.4(b)).

Enforcement:

Technical Specification 5.4.1 requires, in part, that the licensee establish, implement, and maintain applicable procedures recommended in Regulatory Guide 1.33, Revision 2, Appendix A. Section 7 of Appendix A specifies RP procedures for the implementation of ALARA program. Licensee procedures RP-AA-400-1006, Outage Exposure Estimating and Tracking, and RP-AA-401, Operational ALARA Planning and Controls, required effective reviews of the accumulated dose at prescribed intervals to assess the established ALARA controls and to identify/implement prompt dose reduction strategies to minimize exposure to the workers.

Contrary to the above, the licensee failed to implement steps in these procedures to review the accumulated dose at prescribed intervals on multiple occasions during the L2R14 refueling outage.

Corrective actions included instituting appropriate radiological controls and initiating apparent cause evaluations. Because this violation is of very low safety significance and it was entered into the licensees CAP, as AR 1568021, this violation is being treated as an NCV, consistent with Section 2.3.2 of the NRC Enforcement Policy (NCV 05000373/2013005-03; 05000374/2013005-03, Failure to Provide Station ALARA Committee with Necessary Information to Ensure Occupational Radiation Exposure Is Maintained ALARA).

.2 Problem Identification and Resolution (02.06)

a. Inspection Scope

The inspectors evaluated whether problems associated with ALARA planning and controls were being identified by the licensee at an appropriate threshold and were properly addressed for resolution in the licensees CAP.

b. Findings

No findings were identified.

OTHER ACTIVITIES

Cornerstones: Initiating Events, Mitigating Systems, Barrier Integrity, Emergency Preparedness, Public Radiation Safety, Occupational Radiation Safety, and Physical Protection

4OA1 Performance Indicator Verification

.1 Safety System Functional Failures

a. Inspection Scope

The inspectors sampled licensee submittals for the Safety System Functional Failures performance indicator for Units 1 and 2 from the fourth quarter 2012 through the third quarter 2013. To determine the accuracy of the performance indicator (PI) data reported, PI definitions and guidance contained in Nuclear Energy Institute (NEI)document 99-02, Regulatory Assessment Performance Indicator Guideline, Revision 6, dated October 2009, and NUREG-1022, Event Reporting Guidelines 10 CFR 50.72 and 50.73," definitions and guidance were used. The inspectors reviewed the licensees operator narrative logs, operability assessments, maintenance rule records, maintenance WOs, issue reports, event reports, and NRC Integrated Inspection Reports for October 2012 through September 2013 to validate the accuracy of the submittals.

The inspectors also reviewed the licensees issue report database to determine if any problems had been identified with the PI data collected or transmitted for this indicator and none were identified. Documents reviewed are listed in the Attachment to this report.

This inspection constituted two safety system functional failures samples as defined in IP 71151-05.

b. Findings

No findings were identified.

.2 Mitigating Systems Performance Index - Heat Removal System

a. Inspection Scope

The inspectors sampled licensee submittals for the Mitigating Systems Performance Index (MSPI) - heat removal system PI for Units 1 and 2 from the fourth quarter 2012 through the third quarter 2013. To determine the accuracy of the PI data reported, PI definitions and guidance in NEI 99-02 were used. The inspectors reviewed the licensees operator narrative logs, issue reports, event reports, MSPI derivation reports, and NRC Integrated Inspection Reports for October 2012 through September 2013 to validate the accuracy of the submittals. The inspectors reviewed the MSPI component risk coefficient to determine if it had changed by more than 25 percent since the previous inspection, and if so, that the change was in accordance with applicable NEI guidance.

The inspectors also reviewed the licensees issue report database to determine if any problems had been identified with the PI data collected or transmitted for this indicator and none were identified. Documents reviewed are listed in the Attachment to this report.

This inspection constituted two MSPI heat removal system samples as defined in IP 71151-05.

b. Findings

No findings were identified.

4OA2 Problem Identification and Resolution

.1 Routine Review of Items Entered into the Corrective Action Program

a. Inspection Scope

As part of the various baseline inspection procedures discussed in previous sections of this report, the inspectors routinely reviewed issues during baseline inspection activities and plant status reviews to verify they were being entered into the licensees CAP at an appropriate threshold, that adequate attention was being given to timely corrective actions, and that adverse trends were identified and addressed. Attributes reviewed included: identification of the problem was complete and accurate; timeliness was commensurate with the safety significance; evaluation and disposition of performance issues, generic implications, common causes, contributing factors, root causes, extent-of-condition reviews, and previous occurrences reviews were proper and adequate; and that the classification, prioritization, focus, and timeliness of corrective actions were commensurate with safety and sufficient to prevent recurrence of the issue.

Minor issues entered into the licensees CAP as a result of the inspectors observations are included in the Attachment to this report.

These routine reviews for the identification and resolution of problems did not constitute any additional inspection samples. Instead, by procedure they were considered an integral part of the inspections performed during the quarter and documented in Section 1 of this report.

b. Findings

No findings were identified.

.2 Daily Corrective Action Program Reviews

a. Inspection Scope

To assist with the identification of repetitive equipment failures and specific human performance issues for follow-up, the inspectors performed a daily screening of items entered into the licensees CAP. This review was accomplished through inspection of the stations daily condition report packages.

These daily reviews were performed by procedure as part of the inspectors daily plant status monitoring activities and, as such, did not constitute any separate inspection samples.

b. Findings

No findings were identified.

.3 Semi-Annual Trend Review

a. Inspection Scope

The inspectors performed a review of the licensees CAP and associated documents to identify trends that could indicate the existence of a more significant safety issue. The inspectors review was focused on repetitive equipment issues, but also considered the results of daily inspector CAP item screening discussed in Section 4OA2.2 above, licensee trending efforts, and licensee human performance results. The inspectors review nominally considered the 6-month period of July 2013 through December 2013, although some examples expanded beyond those dates where the scope of the trend warranted.

The reviews also included issues documented outside the normal CAP in major equipment problem lists, repetitive and/or rework maintenance lists, departmental problem/challenges lists, system health reports, quality assurance audit/surveillance reports, self-assessment reports, and Maintenance Rule assessments. The inspectors compared and contrasted their results with the results contained in the licensees CAP trending reports. Corrective actions associated with a sample of the issues identified in the licensees trending reports were reviewed for adequacy.

This review constituted one semi-annual trend inspection sample as defined in IP 71152-05.

b. Findings

(1) Failure to Maintain and Report UFSAR Revisions for Safety Analysis and Methodology Changes

Introduction:

The inspectors identified a finding of very low safety significance (Green)and associated Severity Level IV (SL IV) NCV of 10 CFR 50.71 Maintenance of records, making of reports. Specifically, 50.71(e) requires the licensee to periodically update the UFSAR, and by letter dated April 12, 2012, the licensee provided the most recent periodic update; however, certain information contained in the update failed to accurately present changes made since the previous submittal and failed to reflect the current design and licensing basis.

Discussion: By letter dated April 12, 2012, the licensee submitted the latest revision of the UFSAR. The accuracy and completeness was inconsistent with the update requirements of 10 CFR 50.71(e) and the content required to maintain the design and licensing basis associated with Chapter 15 safety analysis. The NRC staff, in its review, utilized Regulatory Guide 1.81, "Content of the Updated Final Safety Analysis Report in Accordance with 10 CFR 50.71(e)," which endorsed NEI 98-03, "Guidelines for Updating Final Safety Analysis Reports," Revision 1, and identified a number of concerns associated with LaSalle facility meeting UFSAR update requirements and the level of detail required pursuant to 10 CFR 50.34(b). Specific examples are provided below.

Example 1: UFSAR, Section 15.A, Cycle Specific Reload Analysis, was revised to state, "This section is retained for its historical information only. The most current safety analyses information can be found in cycle specific transient and reload documents [emphasis added]." This specific UFSAR section does not meet the intent of 10 CFR 50.71(e), to serve as an updated reference document to be used in recurring safety analyses performed by the licensee, the Commission, and other interested parties. The UFSAR is required to capture the current safety analysis and methods of evaluation for the LaSalle units. Relevant information contained in cycle-specific transient and reload documents should either be reflected in the UFSAR text or incorporated by reference into the USFAR. Obsolete information, including methods of evaluation no longer utilized in the current safety analysis, should be removed from the UFSAR. The changes classifying Section 15.A as historic was introduced during the amendment implementation for the LaSalle Measurement Uncertainty Recapture Amendments 198 and 185 in July 2010.

Example 2: The UFSAR states in Section 15.0 that Sections 15.1 through 15.10 are for historic reference; however, the treatment of large sections as historic is overly broad as the LaSalle design and licensing basis continuously evolves and changes over time.

Historic information is generally considered information that does not change, such as initial transient testing, weather and geological information. LaSalle has failed to comprehensively evaluate that all information presented in these Sections meets standards for classification as historic information.

Example 3: UFSAR references are marked as "latest approved revision" without appropriate controls, e.g., General Electric Standard Application for Reactor Fuel, NEDE-24011-P-A, and "Equipment Out-of Service in the Increased Core Flow Domain For LaSalle County Station Units 1 and 2, General Electric document GE-NE-187-62-1191. Formal review of future revision changes is required pursuant to 10 CFR 50.59, Changes, Tests, and Experiments. Adding latest approved revision could allow revision updates to be implemented without performing a 50.59 review.

Title 10 CFR Part 50.59 reviews of methodologies ensure that updated methodologies remain applicable, and used within the approved limitations or conditions.

Example 4: The Core Operating Limits Report (COLR) for the LaSalle units has been incorporated into the UFSAR by reference as part of the Technical Requirements Manual. As such, the COLRs must satisfy 50.71(e) update requirements. The licensee acknowledged in a letter response to request for additional information (RAI) 4(e), dated October 31, 2013, that 12 methodologies listed in the COLR report dated March 6, 2013, were not used for establishing the core operating limits. This condition, where only a subset of COLR methodologies are used, is applicable to the Unit 1 and Unit 2 COLRs provided as part of the UFSAR update letter dated April 12, 2012. The UFSAR update failed to provide sufficient information to allow an understanding of the current design and limiting basis used to develop the current core operating limits. The UFSAR failed to reflect the most recent configuration of the facility within the provisions afforded by 50.71(e) reporting requirements.

Example 5: The April 12, 2012, UFSAR update failed to reflect the LaSalle Unit 2 Cycle 14 specific licensing basis methods utilized for establishing safety analysis and cycle specific reload analysis. LaSalle County Station (LSCS) Unit 2 Cycle 14 contained only Areva Atrium-10 fuel; however, Chapter 15 maintained that both GE and Areva reload methods may be applied to Unit 2. Modifications in place greater than 90 days, such as the change from one type of fuel to another, are generally considered permanent changes to a facility.

Example 6: The UFSAR and Technical Requirements Manual failed to reflect necessary design and licensing basis information. Specifically, LaSalles revised power-to-flow map is not controlled within the UFSAR change process. The power-to-flow map is important to the safe operations of the facility and serves as an input into some safety analysis calculations (e.g., peak containment design pressure calculation). In RAI Response 4(c) from letter dated October 31, 2013, the licensee states that:

The current reload cycle power/flow map for LSCS Unit 1 is in procedure LOA-RR-101, "Unit 1 Reactor Recirculation System Abnormal. The LSCS Unit 2 current reload cycle power/flow map is in procedure LOA-RR-201, "Unit 2 Reactor Recirculation System Abnormal." These power/flow maps are bounded by the information contained in the current Supplemental Reload Licensing Reports (SRLRs), Global Nuclear Fuel (GNF) documents 0000-0136-5291-SRLR, Revision 1 (for LSCS Unit 1) and 0000-0156-1147-SRLR, Revision 1 (for LSCS Unit 2).

The references discussed in the licensees response are not incorporated into the UFSAR. The power-to-flow map is information necessary to understand LaSalles design, safety analysis, and cycle-specific reload analysis. As such, the licensee failed to update the power-to-flow map as required pursuant to 10 CFR 50.71(e) and the level of detail required pursuant to 10 CFR 50.34(b).

Example 7: The UFSAR update failed to remove obsolete methodologies.

Section 15.6.5, Loss-of-Coolant Accidents Resulting from Spectrum of Postulated Piping Breaks within the Reactor Coolant Pressure Boundary, lists a total of four methodologies that may be used for loss-of-coolant accident safety analysis. In an RAI response from October 31, 2013, Engineering Change (EC) 384430, GNF2 Fuel Transition, Revision 3, added a fifth methodology. In review of the LaSalle Units March 2011 and 2012 Annual 10 CFR 50.46 reports, the NRC noted that not all of the methodologies listed in Section 15.6.5 are collectively used in establishing LaSalles licensing basis and analyses of record. The licensee has failed to remove obsolete methodologies from the UFSAR as required by 50.71(e). Specifically, LaSalle does not use SAFE/REFLOOD or EXEM BWR methodologies as the current evaluations models for LaSalle safety analyses. These methodologies have been superseded with updated evaluation models. SAFE/REFLOOD or EXEM BWR evaluations models may not be applicable to future advances in fuel designs.

Example 8: The UFSAR makes general reference in Chapter 15 that current safety analysis is contained in cycle-specific transient and reload licensing documents (e.g. GNF Report SRLR, Supplemental Reload Licensing Report for LaSalle Units).

The licensee has failed to either reflect the current safety analysis results or incorporate the information into the UFSAR by reference. The SRLR report contains the limiting cycle specific safety analysis and operating conditions that are permitted by the core operating limits.

Example 9: The LaSalle UFSAR fails to document methodologies as materials incorporated by reference. The licensee previously deleted Table 1.6.1, Materials incorporated by reference, from the UFSAR. Furthermore, UFSAR Chapter 15 either does not contain materials incorporated by reference or the licensee has not provided reference changes as part of the UFSAR revision pursuant to the 50.71(e) reporting requirement.

Analysis:

The inspectors determined that the licensees failure to maintain the UFSAR as required by 10 CFR Part 50.71(e) was not in accordance with regulations and was a performance deficiency warranting a significance evaluation. Specifically, the NRC staff determined that LSCS UFSAR, Revision 19, did not update information as required by 10 CFR 50.71(e)(1) and (e)(2).

Consistent with the guidance in IMC 0612, Power Reactor Inspection Reports, Appendix B, Issue Screening, dated September 7, 2012, the deficiency was considered to be more than minor, and thus a finding, because if left uncorrected, it could lead to a more significant safety concern. Specifically, methodologies could be utilized without properly performing 50.59 reviews, leading to revisions to operating ranges, such as those found in the power-to-flow map, which may not be reflected in subsequent safety analyses. The failure to update also reduces the ability of the licensee and the NRC to review the stations licensing basis to assess design changes or assess facility safety.

Because the issue had the potential to affect the NRC's ability to perform its regulatory function, the violation was reviewed under the traditional enforcement process.

Consistent with the guidance in Section 6.9, Paragraph d.9, of the NRC Enforcement Policy, the violation associated with this finding was determined to be a SL IV violation.

Additionally, the significance of the underlying technical issue (i.e., failure to reflect the current licensing basis in the UFSAR, failure to remove obsolete methodology for determining design and operating limits, and the failure to supervise how individuals control and modify information presented in LaSalles UFSAR) was evaluated using the SDP. Using IMC 0609, Appendix A, The Significance Determination Process for Findings At-Power, Exhibit 3, dated June 19, 2012, the inspectors reviewed the Barrier Integrity Screening Questions in Appendix A, Exhibit 3, and answered No to all the questions, which screens the issue as very low safety significance (Green).

The finding had a cross-cutting aspect in the area of Human Performance, Work Practices, because the licensee failed to supervise and manage UFSAR changes with LaSalle station staff, the corporate nuclear fuels group, and contractors. Specifically, this was a lapse in departmental supervision and oversight in controlling the content and accuracy of information contained in the licensing and design basis for current safety analysis and core operating limit calculations within the UFSAR. The licensee must ensure supervisory and management oversight of work activities, including contractors, such that nuclear safety is supported (H.4(c)).

Enforcement:

Title 10 CFR Part 50.71(e) requires, in part, that the licensee periodically update the UFSAR originally submitted as part of the application for the operating license, to assure that the information included in the report contains the latest information developed. This submittal shall include the effects of all changes made in a facility or procedures as described in the UFSAR.

Contrary to the above, on April 12, 2012, the licensee failed to update the UFSAR to assure that it contained an accurate description of all necessary UFSAR changes effecting LaSalles current operation, and licensing and design bases.

This failure is characterized as a SL IV violation. This violation is being treated as an NCV in accordance with Section 2.3.2 of the Enforcement Policy, because it was entered into the CAP, as AR 01484305 (NCV 05000373/2013005-04; 05000374/2013005-04, Failure to Maintain and Report UFSAR Revisions for Safety Analysis and Methodology Changes).

As corrective actions, the licensee has captured the issue in the CAP for resolution and plans to review the UFSAR, ensure information is accurately reflected where historical, incorporate the required UFSAR content, and make necessary changes to enhance oversight of future UFSAR revisions, thereby restoring compliance.

(2) Failure to Maintain a Testing Program That Ensured RCIC Components Will Perform Satisfactorily in Service

Introduction:

The inspectors identified a finding of very low safety significance (Green)and associated NCV of 10 CFR Part 50, Appendix B, Criterion XI, Test Control, for the licensees failure to ensure that the surveillance testing program was sufficient to demonstrate safety-related RCIC components of both Units would perform satisfactorily in service.

Description:

The inspectors identified that there were 4 ARs generated from June-September 2013 for the Model 141 M9 U-Cup pressure switch used on the RCIC system. These ARs revealed that the design resulted in a loss of adequate sealing capability along the U-Cup shaft, causing water leakage into the switch housing and subsequent switch fouling. The inspectors then broadened the scope of the research window and noted an apparent trend since September 2010 of problems associated with the function and performance of these pressure switches.

The inspectors performed a historical evaluation and identified that from September 2010 through October 2013 there were 18 instances where these switches failed their calibrations, would not function as designed, and were declared inoperable.

Of those failures, 12 were due to water in the switch housing (Model 141 M9 U-Cup)and 6 were due to a leaky diaphragm (Model 103). There were an additional 2 instances when the switch was degraded due to water and rust/corrosion within the switch housing. Finally, twice in that period, there were spurious isolations of the RCIC system due to Model 141 M9 U-Cup failures.

Despite the repetitive and long-term nature of these switch failures, the licensee failed to resolve the leaking diaphragm in the Model 103 switches and failed to resolve the water intrusion deficiencies of the U-Cup design, resulting in 20 issue reports in a three-year period, due to both of those known deficiencies. On January 12, 2012, AR 1312714 stated that the Model 141 M9 U-Cup had a manufacturing defect allowing water in the switch housing to cause failures. Further, on September 13, 2013, in WO 164997, Instrument Maintenance staff included a statement advising the performance of the switch calibration more frequently. This recommendation was, however, not incorporated in the test program.

The pressure switch calibrations had been performed quarterly in accordance with procedures LIS-RI-101, Unit 1 RCIC Steam Line High Flow Isolation Calibration, Revision 25, and LIS-RI-201, Revision 26, for Unit 2. Per TS Bases B.3.3.6.1, the instruments prevent: 1) a RCIC steam line break from becoming bounding, and the isolation action along with the scram function of the reactor protection system (RPS)ensures that the fuel peak cladding temperature remains below the limits of 10 CFR 50.46. If steam is allowed to continue flowing out of the break, the reactor will depressurize and core uncover can occur; 2) RCIC could also spuriously isolate and not provide core cooling as designed.

Following the inspectors inquiries regarding the adequacy of the surveillance frequency, given the repetitive failures, and regarding the adequacy of the switch design, given the known moisture intrusion issue; as of November 2013, the licensee documented the intentions of increasing the surveillance frequency to every 46 days and expediting the schedule for switch replacement to be completed by April 2014.

Analysis:

The failure to establish an appropriate surveillance frequency for the safety-related switches was not in accordance with 10 CFR 50 Appendix B, Criterion XI, and was a performance deficiency warranting further review. The deficiency was considered more than minor, and thus a finding, in accordance with IMC 0612, Appendix B, Issue Screening, dated September 7, 2012, because it was associated with the Equipment Performance attribute of the Mitigating Systems Cornerstone, and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability, of systems that respond to initiating events to prevent undesirable consequences.

The inspectors determined the finding could be evaluated using the SDP in accordance with IMC 0609, Appendix A, The Significance Determination Process for Findings At-Power, Exhibit 2, dated June 19, 2012. The inspectors reviewed the Mitigating Systems Screening Questions in Appendix A, Exhibit 2, and answered No to all the questions, which screens the issue as Green.

This finding has a cross-cutting aspect in the area of Problem Identification and Resolution, Corrective Action Program, for the failure to take appropriate corrective actions to address adverse trends in a timely manner. Specifically, in the three-year period from September 2010 through October 2013, these switches failed their surveillances 18 times, primarily due to water intrusion into the switch housing, and corrective actions to increase the frequency or address the water intrusion problem were not taken in a timely manner, commensurate with the safety significance of the issue (P.1(d)).

Enforcement:

Title 10 CFR Part 50, Appendix B, Criterion XI, Test Control, requires, in part, that a test program be established to assure that all testing required to demonstrate that structures, systems, and components will perform satisfactorily in service is identified and performed in accordance with written test procedures which incorporate the requirements and acceptance limits contained in applicable design documents.

Contrary to the above, at some point between September 2010 and October 2013, the licensee failed to establish a test program that assured that all testing required to demonstrate that the safety-related RCIC system would perform satisfactorily in service.

Specifically, due to the repetitive failures of the Unit 1 and Unit 2 RCIC Steam Supply Differential Pressure Switches over a three-year period due to a water intrusion-induced failure mechanism, the quarterly frequency at which the switches were tested was incapable of assuring that the SSC would perform satisfactorily in service.

Because this violation was of very low safety significance and it was entered into the licensees CAP (as AR 01562537), this violation is being treated as an NCV, consistent with Section 2.3.2 of the NRC Enforcement Policy (NCV 05000373/2013005-05; 05000374/2013005-05, Failure to Maintain a Testing Program That Ensured RCIC Components Will Perform Satisfactorily in Service).

As corrective actions, the licensee increased the frequency of the switch surveillance testing from every 90 days to every 46 days, and expedited the switch replacement timeline to correct the water intrusion issue in a manner which was more appropriate to the risk significance of the issue.

.4 Annual Sample:

Review of Operator Workarounds

a. Inspection Scope

The inspectors evaluated the licensees implementation of the process used to identify, document, track, and resolve operational challenges. Inspection activities included, but were not limited to, a review of the cumulative effects of the operator workarounds (OWAs) on system availability and the potential for improper operation of the system, for potential impacts on multiple systems, and on the ability of operators to respond to plant transients or accidents.

The inspectors performed a review of the cumulative effects of OWAs. The documents listed in the Attachment to this report were reviewed to accomplish the objectives of the inspection procedure. The inspectors reviewed both current and historical operational challenge records to determine whether the licensee was identifying operator challenges at an appropriate threshold, had entered them into the CAP, and proposed or implemented appropriate and timely corrective actions which addressed each issue.

Reviews were conducted to determine if any operator challenge could increase the possibility of an Initiating Event, if the challenge was contrary to training, required a change from long-standing operational practices, or created the potential for inappropriate compensatory actions. Additionally, all temporary modifications were reviewed to identify any potential effect on the functionality of Mitigating Systems, impaired access to equipment, or required equipment uses for which the equipment was not designed. Daily plant and equipment status logs, degraded instrument logs, and operator aids or tools being used to compensate for material deficiencies were also assessed to identify any potential sources of unidentified operator workarounds.

This review constituted one operator workaround annual inspection sample as defined in IP 71152-05.

b. Findings

No findings were identified.

4OA3 Follow-Up of Events and Notices of Enforcement Discretion

.1 (Closed) Licensee Event Report (LER) 05000374-2012-001-00:

Secondary Containment Inoperable Due to Interlock Doors Open This event occurred on September 18, 2012, while Unit 2 was in Mode 1 at 100 percent reactor power. An equipment operator reported that the Unit 2 reactor building 761 elevation (interlock doors 424 and 314) were open at the same time for approximately 10 seconds. During the time that both interlock doors were open, TS Surveillance Requirement (SR) 3.6.4.1.2, which requires the verification of one secondary containment access door in each access opening is closed, was not met. As a result, TS 3.6.4.1 was entered, and the secondary containment system was declared inoperable until it was verified that one of the doors would remain closed at all times.

This occurrence was reportable under 10 CFR 50.73(a)(2)(v)(C) and 10 CFR 50.73(a)(2)(v)(D) as an event or condition that could have prevented the fulfillment of the safety function of structures or systems that are needed to control the release of radioactive material and to mitigate the consequences of an accident. This event constituted a safety system functional failure.

The inspectors have completed reviewing the adequacy of the licensees implemented and planned corrective actions in response to the events described in the subject LER.

Section 4OA3.3 of this report documents the closure of unresolved item (URI)05000373/2013003-02 and 05000374/2013003-02, which is associated with the review of this LER. Since resolution of this URI was necessary to determine if there were any violations of NRC requirements, this LER may now be closed.

Documents reviewed are listed in the Attachment to this report. This LER is closed.

This event followup constituted one sample as defined in IP 71153-05.

.2 (Closed) LER 05000373/2013-001-00:

Secondary Containment Inoperable Due to Interlock Doors Open This event occurred on February 28, 2013, while Unit 1 was in Mode 5 for refueling outage L2R14. It was reported to the control room that both interlock doors on the Unit 1 reactor building 710 elevation (doors 225 and 226) were open at the same time for approximately 10 seconds. At the time of the event, there were no irradiated fuel movements, core alterations, or operations that could potentially drain the reactor vessel in progress. During the time that both interlock doors were open, TS SR 3.6.4.1.2, which requires the verification of one secondary containment access door in each access opening is closed, was not met. As a result, TS 3.6.4.1 was entered, and the secondary containment system was declared inoperable until it was verified that one of the doors would remain closed at all times.

This occurrence was reportable under 10 CFR 50.73(a)(2)(v)(C) and 10 CFR 50.73(a)(2)(v)(D) as an event or condition that could have prevented the fulfillment of the safety function of structures or systems that are needed to control the release of radioactive material and to mitigate the consequences of an accident. This event constituted a safety system functional failure.

The inspectors have completed reviewing the adequacy of the licensees implemented and planned corrective actions in response to the events described in the subject LER.

Section 4OA3.3 of this report documents the closure of URI 05000373/2013003-02 and 05000374/2013003-02, which is associated with the review of this LER. Since resolution of this URI was necessary to determine if there were any violations of NRC requirements, this LER may now be closed.

Documents reviewed are listed in the Attachment to this report. This LER is closed.

This event followup constituted one sample as defined in IP 71153 05.

.3 (Closed) Unresolved Item 05000373/2013003-02; 05000374/2013003-02:

Potential Failure to Assign Appropriate Safety Classification to Secondary Containment Doors

a. Inspection Scope

A URI associated with secondary containment doors was previously identified by the inspectors. The first aspect of the URI was the potential failure to assign the proper safety classification to the secondary containment doors. The second aspect was associated with the potential failure to provide an adequate test for the secondary containment doors to ensure that the doors will perform satisfactorily while in service.

The Equipment Apparent Cause Evaluation (EACE) 01414490, Unit 2 Reactor Building 761 Interlock Allows Both Doors to be Open, and EACE 01481207, Secondary Containment Door Interlock Failure, both document instances where both doors of an interlock in secondary containment were opened at the same time. This created a condition where secondary containment was declared inoperable and its safety function was considered lost.

b. Findings

Improper Classification of Secondary Containment Doors as Nonsafety-Related

Introduction:

The inspectors identified a finding of very low safety significance (Green)and associated NCV of 10 CFR 50, Appendix B, Criterion II, Quality Assurance Program, for the failure to classify the Unit 1 and 2 secondary containment doors with a quality status of safety-related. The first aspect of the URI was the potential failure to assign the proper safety classification to the secondary containment doors. The second aspect was associated with the potential failure to provide an adequate test for the secondary containment doors to ensure that the doors will perform satisfactorily while in service

Description:

The EACE 1414490, Unit 2 Reactor Building 761 Interlock Allows Both Doors to be Open, and EACE 1481207, Secondary Containment Door Interlock Failure, both document instances where both doors of an interlock in secondary containment were opened at the same time. This created a condition where secondary containment was declared inoperable and its safety function was considered lost.

During the review of the EACEs and associated documentation, the inspectors noted that the secondary containment structure was classified as safety-related but the secondary containment doors were classified as nonsafety-related. Licensee engineering personnel stated that the classification went back to the original licensed design requirements and that the bases were addressed in NUREG 0519, Safety Evaluation Report Related to the Operation of LaSalle County Station Units 1 and 2.

However, following the review of NUREG 0519 and the UFSAR sections associated with classification of structures, systems, and components, and secondary containment functional design, the inspectors determined that there was no distinction between secondary containment and its doors. The design and licensing basis documents reviewed describe secondary containment as a system and did not list exceptions to its components, such as the doors.

The inspectors discussed the issue personnel in the NRC Office of Nuclear Reactor Regulation and determined that the licensee was incorrectly implementing the component classification guidelines and processes.

The inspectors reviewed licensee procedure CC-AA-304, Component Classification, Revision 5. Procedure CC-AA-304, Section 4.1.1, stated:

If the component performs any safety related function then Enter SR in the Safety Class Field. Considerations listed below3. Items for which credit has been taken in the Accident Analyses to respond to the Design Basis Accidents and Transients to meet the basic Safety Related functions of the general criteria above.

In response to the inspectors questions, the licensee stated that the secondary containment dose assessment (Design Analyses L-003067 and L003068) credits the doors being in their normal closed position and that the only effluent release path credited was through Standby Gas Treatment System.

Per Chapter 15 of the UFSAR, L-003068 was the Accident Analysis for a loss of coolant accident. Therefore, the inspectors concluded that per the licensees procedure for component classification the secondary containment doors are safety-related.

For the second part of the URI, the inspectors reviewed LOS-CS-M1, Secondary Containment Integrity. This procedure outlined the steps to test the secondary containment doors as required per TS SR to verify one secondary containment access door in each access opening is closed. The frequency stated for this TS SR is 31 days.

The inspectors noted that prior to December 2011, the secondary containment doors were tested by opening one of the two interlock doors and physically challenging the other one to verify that the interlock mechanism had actuated and that the two doors could not be opened at the same time. However, in January 2012, operations personnel changed LOS-CS-M1 such that only the light indication would be checked, and the doors would not be physically challenged. The licensee stated that the reason for this change was to prevent inadvertent losses of the secondary containment function during the testing. The inspectors noted that merely checking the light to verify the interlock had engaged when one of the doors was open would not completely ensure that the doors would perform satisfactorily while in service. Specifically, since the light indication and the door lock were wired in parallel, a test could be determined satisfactory (based on the light indication) when the locking mechanism could be non-functional.

However, upon further review of TS SR 3.6.4.1.2 which stated, Verify one secondary containment access door in each access opening is closed, the inspectors concluded that this surveillance requirement did not require a verification of the door interlock.

Although the inspectors felt the verification of the interlock was prudent, a review of SR 3.6.1.2.2 for the primary containment air lock very specifically requires the verification of the interlock by stating, Verify only one door in the primary containment air lock can be opened at a time.

Analysis:

The inspectors determined that the failure to classify the secondary containment doors as safety-related was contrary to CC-AA-304, Component Classification, Revision 5, and was a performance deficiency.

The deficiency was considered more than minor, and thus a finding, in accordance with IMC 0612, Appendix B, Issue Screening, dated September 7, 2012, because if left uncorrected, it would become a more significant safety concern. Specifically, by removing the quality assurance requirements for this part, the licensee reduced the assurance that replacement parts were of sufficient quality to assure reliable service during and following design basis events. The inspectors concluded this finding was associated with the Barrier Systems Cornerstone.

The inspectors determined the finding could be evaluated using the SDP in accordance with IMC 0609, Significance Determination Process, Appendix A, Exhibit 3, Barrier Integrity Screening Questions, dated June 19, 2012. The inspectors answered both question in Section C, No, therefore the finding screened as of very low safety significance (Green) because the finding was a qualification deficiency confirmed not to result in loss of operability or functionality.

The inspectors did not identify a cross-cutting aspect associated with this finding, primarily because the original classification occurred more than three years ago and is not representative of current performance.

Enforcement:

Title 10 CFR Part 50, Appendix B, Criterion II, Quality Assurance Program, requires, in part, that the licensee identify the structures, systems, and components to be covered by the quality assurance program. Licensee procedure CC AA 304, Component Classification, Revision 5, Section 4.1.1, required that equipment that was taken credit for in a safety analysis be safety-related. The secondary containment doors were credited in the licensees loss-of-coolant-accident analysis.

Contrary to the above, until September 2013, the licensee failed to assure that Units 1 and 2 secondary containment doors were identified as components to be covered by the quality assurance program. Specifically, the licensee inappropriately classified the Units 1 and 2 secondary containment doors as nonsafety-related. Because this violation was of very low safety significance and it was entered into the licensees CAP (as AR 01557738), this violation is being treated as an NCV, consistent with Section 2.3.2 of the NRC Enforcement Policy (NCV 05000373/2013005-06; 05000374/2013005-06, Improper Classification of Secondary Containment Doors as Nonsafety-Related).

As corrective actions, the licensee was upgrading the affected doors to the proper safety classification. Additionally, the licensee is in the process of installing an upgraded, more reliable, interlock system.

Unresolved Item 05000373/2013003-02; 05000374/2013003-02, Potential Failure to Assign Appropriate Safety Classification to Secondary Containment Doors, is closed.

4OA6 Management Meetings

Exit Meeting Summary

On January 8, 2014, the inspectors presented the inspection results to Mr. P. Karaba, Site Vice-President, and other members of the licensee management and staff. The licensee acknowledged the issues presented. The inspectors confirmed that none of the potential report input discussed was considered proprietary.

.2 Interim Exit Meetings

Interim exits were conducted for:

• inspection results for the areas of radiological hazard assessment and exposure controls; and occupational ALARA planning and controls with Mr. P. Karaba, Site Vice-President, on October 4, 2013; and

• inspection results from the annual and biennial licensed operator requalification program assessment with Mr. H. Vinyard, Plant Manager, on November 22, 2013.

The inspectors confirmed that none of the potential report input discussed was considered proprietary. Proprietary material received during the inspection was returned to the licensee.

ATTACHMENT:

SUPPLEMENTAL INFORMATION

KEY POINTS OF CONTACT

Licensee

P. Karaba, Site Vice-President

H. Vinyard, Plant Manager

J. Kowalski, Engineering Manager

B. Maze, Project Management

A. Schierer, Engineering Programs

K. Hall, Buried Piping Program Owner

V. Chopra, Engineering Programs

G. Ford, Regulatory Affairs Manager

L. Blunk, Regulatory Affairs

S. Shields, Regulatory Affairs

J. Vergara, Regulatory Assurance

B. Hilton, Design Manager

J. Houston, Nuclear Oversight Manager

L. Ekern, Nuclear Oversight

D. Amezaga, Design Engineer

J. Bendis, Engineer

J. Hughes, Emergency Preparedness Coordinator

J. Shields, Invessel Visual Inspection Program Supervisor

S. Tanton, Engineer

A. Daniels, Exelon Emergency Preparedness Manager

M. Hayworth, Emergency Preparedness Manager

S. Tutoky, Senior Chemist

M. Martin, Chemistry Developmental Manager

T. Halliday, Radiation Protection Operation Manager

J. Mosher, Radiation Protection Manager

C. Howard, Radiation Protection Operation Manager

S. Koval, Radwaste Shipping Specialist

A. Baker, Dosimetry Specialist

J. Bauer, Training Director

T. Dean, Operations Training Manager

Nuclear Regulatory Commission

M. Kunowski, Chief, Reactor Projects Branch 5

LIST OF ITEMS

Opened

05000373/2013005-01 NCV Failure to Follow Winter Operation Preparation Procedure (1R01)

05000373/2013005-02;

05000374/2013005-02 NCV Failure to Comply with Radiation Work Permit Requirements (2RS1)

05000373/2013005-03;

05000374/2013005-03 NCV Failure to Provide Station ALARA Committee With Necessary Information to Ensure Occupational Radiation Exposure Is Maintained ALARA (2RS2)

05000373/2013005-04;

05000374/2013005-04 NCV Failure to Maintain and Report UFSAR Revisions for Safety Analysis and Methodology Changes (4OA2.3(1))

05000373/2013005-05;

05000374/2013005-05 NCV Failure to Maintain a Testing Program That Ensured RCIC Components Will Perform Satisfactorily in Service (4OA2.3(2))

05000373/2013005-06;

05000374/2013005-06 NCV Improper Classification of Secondary Containment Doors as Nonsafety-Related (4OA3.3)

Closed

05000373/2013005-01 NCV Failure to Follow Winter Operation Preparation Procedure (1R01)

05000373/2013005-02;

05000374/2013005-02 NCV Failure to Comply with Radiation Work Permit Requirements (2RS1)

05000373/2013005-03;

05000374/2013005-03 NCV Failure to Provide Station ALARA Committee With Necessary Information to Ensure Occupational Radiation Exposure Is Maintained ALARA (2RS2)

05000373/2013005-04;

05000374/2013005-04 NCV Failure to Maintain and Report UFSAR Revisions for Safety Analysis and Methodology Changes (4OA2.3(1))

05000373/2013005-05;

05000374/2013005-05 NCV Failure to Maintain a Testing Program that Ensured RCIC Components Will Perform Satisfactorily in Service (4OA2.3(2))

05000374/2012-001-00 LER Secondary Containment Inoperable Due to Interlock Doors Open (4OA3.1)

05000373/2013-001-00 LER Secondary Containment Inoperable Due to Interlock Doors Open (4OA3.2)

05000373/2013003-02;

05000374/2013003-02 URI Potential Failure to Assign Appropriate Safety Classification to Secondary Containment Doors (4OA3.3)

05000373/2013005-06;

05000374/2013005-06 NCV Improper Classification of Secondary Containment Doors as Nonsafety-Related (4OA3.3)