IR 05000237/2016004

NRC Integrated Inspection Report 05000237/2016004; 05000249/2016004; 05000237/2016501 and 05000249/2016501

ML17030A207
Person / Time
Site:	Dresden
Issue date:	01/30/2017
From:	Jamnes Cameron Reactor Projects Region 3 Branch 4
To:	Bryan Hanson Exelon Generation Co, Exelon Nuclear
References
EA-13-068 IR 2016004, IR 2016501
Download: ML17030A207 (56)

Text

UNITED STATES NUCLEAR REGULATORY COMMISSION REGION III 2443 WARRENVILLE RD. SUITE 210 LISLE, IL 60532-4352 January 30, 2017

EA-13-068 Mr. Bryan Senior VP, Exelon Generation Co., LLC President and CNO, Exelon Nuclear 4300 Winfield Road

Warrenville, IL 60555 SUBJECT: DRESDEN NUCLEAR POWER STATION, UNITS 2 AND 3-NRC INTEGRATED INSPECTION REPORT 05000237/2016004; 05000249/2016004; 05000237/2016501 AND 05000249/2016501

Dear Mr. Hanson:

On December 31, 2016, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at your Dresden Nuclear Power Station, Units 2 and 3. On January 13, 2017, the NRC inspectors discussed the results of this inspection with Mr. J. Washko and other members of your staff. The inspectors documented the results of this inspection in the enclosed inspection report. The NRC also completed its annual inspection of the Emergency Preparedness Program. This inspection began on January 1, 2016, and issuance of this letter closes Inspection Report Number 2016501.

This inspection confirmed your implementation of the Confirmatory Order issued to you by the NRC on October 28, 2013 and updated on May 4, 2015. The inspectors independently reviewed information you provided, inspected records of activities that were completed, and determined that your actions were in compliance with the requirements delineated in the Confirmatory Order. The NRC has no further questions on this issue. There were no findings in this area.

Based on the results of this inspection, no U.S. Nuclear Regulatory Commission (NRC) identified findings and one self-revealed finding of very-low safety significance (Green) was identified. The finding was determined to involve a violation of NRC requirements. Further, the inspectors documented a licensee-identified violation which was determined to be of very low safety significance (Green) in this report. However, because of the very-low safety significance and because the issues were entered into your Corrective Action Program (CAP), the NRC is treating the issues as Non-Cited Violations (NCVs), in accordance with Section 2.3.2 of the

NRC's Enforcement Policy. If you contest the violations or significance of the NCVs, you should provide a response within 30 days of the date of this inspection report, with the basis for your denial, to the U.S. Nuclear Regulatory Commission, ATTN: Document Control Desk, Washington, DC 20555-0001, with copies to: (1) the Regional Administrator, Region III; (2) the Director, Office of Enforcement, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001; and (3) the NRC Resident Inspector at the Point Beach Nuclear Plant. In addition, if you disagree with the cross-cutting aspect assignment to any finding in this report, you should provide a response within 30 days of the date of this inspection report, with the basis for your disagreement, to the Regional Administrator, Region III, and the NRC Resident Inspector at the Dresden Nuclear Power Station. In accordance with 10 CFR 2.390 of the NRC's "Rules of Practice," a copy of this letter, its enclosure(s), and your response, (if any), will be made available electronically for public inspection in the NRC Public Document Room or from the NRC's Agencywide Documents Access and Management System (ADAMS), accessible from the NRC Web site at http://www.nrc.gov/reading-rm/adams.html. To the extent possible, your response should not include any personal privacy or proprietary, information so that it can be made available to the Public without redaction.

Sincerely,

/RA/

Jamnes Cameron, Chief Branch 4 Division of Reactor Projects

Docket Nos: 50-237; 50-249 License Nos: DPR-19; DPR-25

Enclosure:

IR 05000237/2016004; 05000249/2016004; 05000237/2016501; 05000249/2016501

REGION III Docket Nos: 50-237; 50-249 License Nos: DPR-19; DPR-25 Report No: 05000237/2016004; 05000249/2016004; 05000237/2016501; 05000249/2016501 Licensee: Exelon Generation Company, LLC Facility: Dresden Nuclear Power Station, Units 2 and 3 Location: Morris, IL Dates: October 1 through December 31, 2016 Inspectors: G. Roach, Senior Resident Inspector R. Elliott, Resident Inspector M. Domke, Inspector, Region III Office G. Edwards, Health Physicist M. Garza, Emergency Preparedness Inspector T. Go, Health Physicist M. Holmberg, Reactor Inspector J. Maynen, Senior Physical Security Inspector L. Torres, ASME Inspector, Illinois Emergency Management Agency Approved by: J. Cameron, Chief Branch 4 Division of Reactor Projects

SUMMARY

Inspection Report 05000237/2016004, 05000249/2016004; 10/01/2016-12/31/2016; 05000237/2016501, 05000249/2016501; 01012016 - 12/30/2016; Dresden Nuclear Power Station, Units 2 & 3; Radiological Hazard Assessment and Exposure Controls. This report covers a 3-month period of inspection by resident inspectors and announced baseline inspections by regional inspectors and the annual review of emergency preparedness.

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 April 29, 2015. Cross-cutting aspects are determined using IMC 0310, "Aspects Within the Cross-Cutting Areas," dated December 4, 2014. All violations of NRC requirements are dispositioned in accordance with the NRC's Enforcement Policy, dated November 1, 2016. The NRC's program for overseeing the safe operation of commercial nuclear power reactors is described in NUREG-1649, "Reactor Oversight Process," Revision 6, dated July 2016.

Cornerstone: Occupational Radiation Safety

Green.

A finding of very-low safety significance, and an associated Non-Cited Violation (NCV) of Technical Specification 5.4.1 was self-revealed when workers violated a radiation work permit (RWP) by entering an area that was outside of the scope of the original RWP brief without obtaining a required appropriate brief, resulting in these workers receiving unplanned electronic dosimeter dose rate alarms. These workers immediately exited the area and reported the event to the radiation protection staff. The licensee entered these issues as two separate events into their CAP as Issue Reports (IR) 02735594 and IR 02735651. The inspectors determined that the performance deficiency was more than minor in accordance with Inspection Manual Chapter 0612, Appendix B, because the finding impacted 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. Specifically, worker entry into areas beyond the RWP briefing could lead to unintended dose. The finding was determined to be of very-low safety significance (Green) in accordance with Inspection Manual Chapter 0609, Appendix C, "Occupational Radiation Safety Significance Determination Process," dated August 19, 2008, because: (1) it did not involve as-low-as-reasonably-achievable planning or work controls, (2) there was no overexposure, (3) there was no substantial potential for an overexposure, and (4) the ability to assess dose was not compromised. The inspectors concluded that the cause of the finding involved a cross-cutting component in the human performance area of challenging the unknown because the individual did not stop when faced with an uncertain condition. Risks were not evaluated and managed before proceeding [H.11]. (Section 2RS1.3)

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Cornerstone: Emergency Preparedness One violation of very low safety significance (

Green), which was identified by the licensee, has been reviewed by the inspector. Corrective actions taken, or planned, by the licensee have been entered into the licensee's Corrective Action Program. This violation and corrective actions are listed in Section 4OA7 of this report.

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REPORT DETAILS

Summary of Plant Status Unit 2 operated at or near full power from the start of the inspection period until December 10 th , when operators reduced power below 20 percent to perform a planned maintenance outage D2M18. The unit returned to near full power on December 11 th, where it operated at or near for the remainder of the inspection period. Unit 3 began the inspection period in coast down to refueling outage D3R24 at 82 percent reactor power (797 MWe). Operators briefing reduced Unit output to 640 MWe on October 20 th , to conduct Isolation Condenser testing. Unit shutdown began on the evening of October 30 th , with the unit coming offline at midnight, beginning D3R24. Unit startup commenced early on November 16 th, and the unit was synchronized to the grid later that night, ending the outage. The Unit returned to full power conditions on November 18 th, where the unit operated at or near for the remainder of the inspection period.

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 reviewed the licensee's preparations for winter conditions to verify that the plant's design features and implementation of procedures were sufficient to protect===

mitigating systems from the effects of adv erse 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 licensee's procedures used to mitigate or respond to adverse weather conditions. Additionally, the inspectors reviewed the Updated Final Safety Analysis Report (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 CAP items to verify that the licensee was identifying adverse weather issues at an appropriate threshold and entering them into their 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:

• condensate storage tanks;

• cribhouse; and

• FLEX buildings.

5 This inspection constituted one winter seasonal readiness preparations sample as defined in IP 71111.01-05.

b. Findings

No findings were identified.

1R04 Equipment Alignment

.1 Quarterly Partial System Walkdowns

a. Inspection Scope

The inspectors performed partial system walkdowns of the following risk-significant systems:

• 3B core spray while Buss 33-1 out-of-service(OOS);

• 2A turbine building closed cooling water (TBCCW) with 2B TBCCW OOS; and

• unit 2 emergency diesel generator (EDG) with 2/3 EDG OOS. The inspectors selected these systems based on their risk significance relative to the Reactor Safety Cornerstones at the time they were 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, Technical Specification (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 systems incapable of performing their intended functions. The inspectors also walked down accessible portions of the systems to verify system components and support equipment were aligned correctly and operable. The inspectors examined the material condition of the components and observed operating parameter s 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 three partial system walkdown samples 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 which were focused on availability, accessibility, and the condition of firefighting equipment in the following risk-significant plant areas:

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• Fire Zone 8.2.5D, U3 low pressure heater bay elevation 517';

• Fire Zone 8.2.5E, U3 high pressure heaters/steam line elevation 517';

• Fire Zone 1.1.2.2, U2 reactor building ground floor elevation 517'; and

• Fire Zone 11.1.1, U3 southwest corner room elevation 476'. 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 licensee's fire plan.

The inspectors selected fire areas based on their overall contribution to internal fire risk as documented in the plant's 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 plant's 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 licensee's 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.

1R08 Inservice Inspection Activities

From October 31, 2016, through November 8, 2016, the inspectors conducted a review of the implementation of the licensee's Inservice Inspection (ISI) Program for monitoring degradation of the reactor coolant system, risk significant piping and components and containment systems in Unit 3. The ISIs described in Sections 1R08.1 and 1R08.5 below constituted one inspection sample as defined in Inspection Procedure 71111.08-05.

.1 Piping Systems Inservice Inspection

a. Inspection Scope

The inspectors observed the following Non-Destructive Examinations mandated by the American Society of Mechanical Engineers (ASME)Section XI Code to evaluate compliance with the ASME Code Section XI and Section V requirements and if any indications and defects were detected, to determine if these were dispositioned in accordance with the ASME Code or an U.S. Nuclear Regulatory Commission (NRC)approved alternative requirement:

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• manual ultrasonic (UT) examination of the core spray system tee-to-pipe welds 3/2/1404-10/10-48 and 10-48.1;

• manual UT examination of the high pressure coolant injection system elbow to pipe weld 3/2/2306-24/24-4;

• manual UT examination of the main steam system elbow-to-pipe welds 3/1/3001C-20/20-K2 and 3/1/3001D-20/20-K2;

• magnetic particle examination of the main steam system support attachment welds 3/1/3001B-20/M-564K SHT 25; and

• magnetic particle examination of the reactor pressure vessel upper head-to-flange weld 3/1/RPV UPP HD/-THD-FLG, RPV. The inspectors observed the following examinations conducted as part of the licensee's commitments to NRC Generic Letter 88-01. "NRC Position on Intergranular Stress Corrosion Cracking in Boiling Water Reactor (BWR) Austenitic Piping," and BWRVIP 75a, "BWR Vessel and Internals Project Technical Basis for Revisions to Generic Letter 88-01 Inspection Schedules," to determine if the examinations were conducted in accordance with the licensee's Augmented Inspection Program, industry guidance documents and associated licensee examination procedures and if any indications and defects were detected, to determine if these were dispositioned in accordance with approved procedures and NRC requirements:

• automated, phased array UT examination of the isolation condenser system safe end-to-nozzle welds 3/1/1302-14/N5B-3 and 3/2/1302A-12/12-7. During the prior outage nondestructive surface and volumetric examinations, the licensee did not identify any relevant/recordable indications. Therefore, no NRC review was completed for this inspection procedure attribute. The inspectors reviewed records of the following pressure boundary weld completed for a risk significant system since the last Unit 3 refueling outage to determine if the licensee applied the preservice Non-Destructive Examinations and acceptance criteria required by the construction Code, and/or the NRC approved Code relief request:

• Repair Needed for Reactor Pressure Vessel Head Spray Pipe Flange - Remove Groove Indications, Weld 01 (Work Order 01595080-01).

Additionally, the inspectors reviewed the welding procedure specification and supporting weld procedure qualification records to determine whether the weld procedures were qualified in accordance with the requirements of the Construction Code and the ASME Code,Section IX.

b. Findings

No findings were identified.

.5 Identification and Resolution of Problems

a. Inspection Scope

The inspectors performed a review of ISI-related problems entered into the licensee's Corrective Action Program and conducted interviews with licensee staff to determine if:

• the licensee had established an appropriate threshold for identifying ISI-related problems;

• the licensee had performed a root cause (if applicable) and taken appropriate corrective actions; and

• the licensee had evaluated operating experience and industry generic issues related to ISI and pressure boundary integrity. The inspectors performed these reviews to evaluate compliance with Title 10 of the Code of Federal Regulations, Part 50, Appendix B, Criterion XVI, "Corrective Action," requirements. The corrective action documents reviewed by the inspectors are listed in the Attachment to this report.

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 11, 2016, the inspectors observed a crew of licensed operators in the plant's simulator during licensed operator requalification training. The inspectors verified that operator performance was adequate, evaluators were identifying and documenting crew performance problems, and that training was being conducted in accordance with licensee procedures. The inspectors evaluated the following areas:

• licensed operator performance;

• crew's 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.

9 The crew's 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 71111.11-05.

b. Findings

No findings were identified.

.2 Resident Inspector Quarterly Observation During Periods of Heightened Activity or Risk

(71111.11Q)

a. Inspection Scope

On December 10, 2016, the inspectors observed the Unit 2 power reduction and drywell de-inerting to address steam leaks on moisture separator piping as well as uncoupled drywell equipment and floor drain sump pumps during D2M18

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

• licensed operator performance;

• crew's 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. 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 71111.11-05.

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 system and the overall Maintenance Rule program:

• Unit 3 shutdown cooling system; and 10

• the inspectors assessed the licensee's overall Maintenance Rule program health by reviewing the licensee's 10 CFR 50.65(a)(3) report covering October 1, 2014, through September 30, 2016. 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 activities affecting risk-significant and safety-related equipment listed below to verify that the appropriate risk assessments were performed prior to removing equipment for work:

• Unit 2 YELLOW online risk with the low pressure coolant injection (LPCI) swing bus relay OOS;

• Unit 3 YELLOW shutdown risk due to lowered inventory;

• Unit 2 YELLOW online risk with Unit 3 250 Vdc battery OOS;

• Unit 3 YELLOW shutdown risk during Bus 33-1 outage; and

• Unit 3 YELLOW online risk with high pressure coolant injection (HPCI) OOS.

11 These activities were selected based on their potential risk significance relative to the Reactor Safety Cornerstones. As applicable for each 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 unit supervisor, 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 during this inspection are listed in the Attachment to this report.

These maintenance risk assessments and emergent work control activities constituted five samples 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:

• potential pre-conditioning of Unit 2 LPCI swing bus relay testing;

• Unit 2 HPCI Group IV containment isolation relay failure;

• Operating Experience (OPEX): electr omagnetic relief valve (ERV) lock washer verification;

• 1C/1D MSIV closure time out of tolerance;

• Unit 2, 2-1601-33F, torus to drywell vacuum breaker would not close properly; and

• horizontal seismic stabilizing bar loose affecting 10 Unit 3 control rod drive insert lines and drywell penetration X-139A. 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 TS and UFSAR to the licensee's 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 corrective action 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.

12 This operability inspection constituted six 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 following permanent modifications:

• Unit 3 essential service system (ESS) uninterruptible power supply (UPS) safety-related power supply bypass line installation; and

• Unit 3 main steam line isolation valve (MSIV) limit switch relocation. The inspectors reviewed the configuration 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, 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 Attachment to this report. This inspection constituted two permanent plant modification samples 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 (PM) activities to verify that procedures and test activities were adequate to ensure system operability and functional capability:

• 3A reactor recirculation pump seal package following replacement; 13

• Unit 3 drywell equipment drain sump (DWEDS) primary containment isolation valve (PCIV), 3-2001-5 valve timing, local leak rate test (LLRT), and leakage test post replacement;

• Unit 3 inboard 'D' MSIV following internals replacement;

• Unit 3 ERVs following actuator replacement and refurbishment; and

• Unit 3 HPCI following motor gear unit and motor speed changer replacements. 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 corrective action 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.

1R20 Outage Activities

.1 Refueling Outage Activities

a. Inspection Scope

The inspectors reviewed the Outage Safety Plan (OSP) and contingency plans for the Unit 3 refueling outage (RFO) D3R24, conducted October 31 - November 16, 2016, to confirm that the licensee had appropriately considered risk, industry experience, and previous site-specific problems in developing and implementing a plan that assured maintenance of defense-in-depth. During the RFO, the inspectors observed portions of the shutdown and cooldown processes and monitored licensee controls over the outage activities listed below:

• licensee configuration management, including maintenance of defense-in-depth commensurate with the OSP for key safety functions and compliance with the applicable TS when taking equipment out of service;

• implementation of clearance activities and confirmation that tags were properly hung and equipment appropriately configured to safely support the work or testing; 14

• installation and configuration of reactor coolant pressure, level, and temperature instruments to provide accurate indication, accounting for instrument error;

• controls over the status and configuration of electrical systems to ensure that TS and OSP requirements were met, and controls over switchyard activities;

• monitoring of decay heat removal processes, systems, and components;

• controls to ensure that outage work was not impacting the ability of the operators to operate the spent fuel pool cooling system;

• reactor water inventory controls including flow paths, configurations, and alternative means for inventory addition, and controls to prevent inventory loss;

• controls over activities that could affect reactivity;

• maintenance of secondary containment as required by TS;

• licensee fatigue management, as required by 10 CFR 26, Subpart I;

• refueling activities, including fuel handling and core verification;

• startup and ascension to full power operation, tracking of startup prerequisites, walkdown of the drywell and torus (primary containment) to verify that debris had not been left which could block emergency core cooling system suction strainers, and reactor physics testing; and

• licensee identification and resolution of problems related to RFO activities.

Documents reviewed are listed in the Attachment to this report. This inspection constituted one RFO sample as defined in IP 71111.20-05.

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 3 MSIV LLRT (isolation valve);

• Unit 3 feed water check valves LLRT (isolation valve);

• Unit 3 Division I emergency core cooling system (ECCS) functional and under voltage test (routine); and

• Unit 3 ATWS [anticipated transient without scram] RPT [recirculation pump trip] (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; 15

• 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, 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;

• 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, and two containment isolation valve samples as defined in IP 71111.22, Sections-02 and-05. In addition, the inspectors did not identify any performance degradation in the RCS leakage for the entire cycle. The reactor coolant system leak detection inspection sample was not performed as defined in IP 71111.22, Section-02.

b. Findings

No findings were identified.

1EP3 Emergency Response Organization Staffing and Augmentation System

.1 Emergency Response Organization Staffing and Augmentation System

a. Inspection Scope

In Dresden Nuclear Power Station, Units 2 and 3 - NRC Integrated Inspection Report 05000237/2016001 and 05000249/2016001 (ADAMS Accession Number ML16120A618), the inspectors documented the completion of a partial sample for Inspection Procedure 71114.03, Emergency Response Organization Staffing and Augmentation System, and committed to completing a full sample by the end of the

16 calendar year of 2016. This inspection procedure was completed through an in-office review of the licensee's procedures and corrective actions associated with the site's backup method of Emergency Response Organization (ERO) activation and

augmentation. The completion of this ERO augmentation testing inspection constituted one sample as defined in Inspection Procedure (IP) 71114.03.

b. Findings

No findings were identified.

1EP4 Emergency Action Level and Emergency Plan Changes

a. Inspection Scope

The regional inspector performed an in-office review of the latest revisions to the Emergency Plan, Emergency Action Levels (EALs), and EAL Bases document to determine if these changes decreased the effectiveness of the Emergency Plan.

The inspector also performed a review of the licensee's Title 10, Code of Federal Regulations, Part 50.54(q) change process, and Emergency Plan change documentation to ensure proper implementation for maintaining Emergency Plan integrity. 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 specific documents reviewed during this inspection are listed in the Attachment to this report. This EAL and Emergency Plan Change inspection constituted one sample as defined in IP 71114.04.

b. Findings

No findings were identified.

1EP6 Drill Evaluation

.1 Training Observation

a. Inspection Scope

The inspector observed a simulator training evolution for licensed operators on December 13, 2016, which required emergency plan implementation by a licensee operations crew. This evolution was planned to be evaluated and included in performance indicator data regarding drill and exercise performance. The inspectors observed event classification and notification activities performed by the crew. The inspectors also attended the post-evolution critique for the scenario. The focus of the inspectors' activities was to note any weaknesses and deficiencies in the crew's performance and ensure that the licensee evaluators noted the same issues and entered them into the corrective action program. As part of the inspection, the inspectors reviewed the scenario package and other documents listed in the Attachment to this report.

17 This inspection of the licensee's training evolution with emergency preparedness drill aspects constituted one sample as defined in IP 71114.06-06.

b. Findings

No findings were identified.

2RS1 Radiological Hazard Assessment and Exposure Controls

.1 Radiological Hazard Assessment (02.02)

a. Inspection Scope

The inspectors assessed the licensee's current and historic isotopic mix, including alpha emitters and other hard-to-detect radionuclides. The inspectors evaluated whether survey protocols were reasonable to identify the magnitude and extent of the radiological hazards. The inspectors determined if there have been changes to plant operations since the last inspection that may have resulted in a significant new radiological hazard for onsite individuals. The inspectors evaluated whether the licensee assessed the potential impact of these changes and implemented periodic monitoring, as appropriate, to detect and quantify the radiological hazard. The inspectors reviewed the last two radiological surveys from selected plant areas and evaluated whether the thoroughness and frequency of the surveys were appropriate for the given radiological hazard. The inspectors conducted walkdowns of the facility, including radioactive waste processing, storage, and handling areas to evaluate material conditions and performed independent radiation measurements as needed to verify conditions were consistent with documented radiation surveys. The inspectors assessed the adequacy of pre-work surveys for select radiologically risk-significant work activities. The inspectors evaluated the Radiological Survey Program to determine if hazards were properly identified. The inspectors discussed procedures, equipment, and performance of surveys with radiation protection staff and assessed whether technicians were knowledgeable about when and how to survey areas for various types of radiological hazards. The inspectors observed work in potential airborne areas to assess whether air samples were being taken appropriately for their intended purpose and reviewed various survey records to assess whether the samples were collected and analyzed appropriately. The inspectors also reviewed the licensee's program for monitoring contamination which has the potential to become airborne. These inspection activities constituted one complete sample as defined in Inspection Procedure (IP) 71124.01-05.

b. Findings

No findings were identified.

.2 Radiological Hazards Control and Work Coverage (02.05)

a. Inspection Scope

The inspectors evaluated ambient radiological conditions during tours of the facility. The inspectors assessed whether the conditions were consistent with applicable posted surveys, radiation work permits (RWPs), and worker briefings. The inspectors evaluated the adequacy of radiological controls, such as required surveys, radiation protection job coverage, and contamination controls. The inspectors evaluated the licensee's use of electronic alarming dosimeters in high noise areas as high radiation area monitoring devices. The inspectors assessed whether radiation monitoring devices were placed on the individual's 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 a U.S. Nuclear Regulatory Commission approved method of determining effective dose equivalent. The inspectors reviewed the application of dosimetry to effectively monitor exposure to personnel in work areas with significant dose rate gradients. For select airborne area RWPs, the inspectors reviewed airborne radioactivity controls and monitoring, the potential for significant airborne levels, containment barrier integrity, and temporary filtered ventilation system operation. The inspectors examined the licensee's physical and programmatic controls for highly activated or contaminated materials stored within pools and assessed whether appropriate controls were in place to preclude inadvertent removal of these materials from the pool. These inspection activities constituted one complete sample as defined in IP 71124.01-05.

b. Findings

No findings were identified.

.3 Radiation Worker Performance and Radiation Protection Technician Proficiency (02.07)

a. Inspection Scope

The inspectors observed radiation worker performance and assessed their performance with respect to radiation protection work requirements, the level of radiological hazards present, and RWP controls. The inspectors assessed worker awareness of electronic alarming dosimeter set points, stay times, or permissible dose for radiologically significant work as well as expected response to alarms.

19 The inspectors observed radiation protection technician performance and assessed whether the technicians were aware of the radiological conditions and RWP controls and whether their performance was consistent with training and qualifications for the given radiological hazards. The inspectors observed radiation protection technician performance of radiation surveys and assessed the appropriateness of the instruments being used, including calibration and source checks. These inspection activities constituted one complete sample as defined in IP 71124.01-05.

b. Findings

Introduction:

A self-revealed finding of very-low safety significance (Green) and associated NCV of Technical Specification 5.4.1, Procedures, was identified when radiation workers violated RWP requirements by entering areas that were outside of the scope of the received RWP briefings. The workers' failure to comply with the requirements to obtain a briefing for actions that were beyond the scope in the received RWP briefing were within the licensee's ability to foresee and correct and should have been prevented, therefore constituting a performance deficiency. This action resulted in workers receiving unplanned electronic dosimeter dose rate alarms.

Description:

On October 31, 2016, the radiation protection staff gave RWP briefings to two individuals on two separate outage activities, specifically;

(1) inside Unit 3 turbine building 517 feet elevation performing feedwater pump maintenance; and

(2) for work inside the moisture separator area installing scaffold decking. The two workers worked under separate RWPs; the first was RWP DR-03-16-00816 and the second was RWP DR-03-16-00411. Both RWPs specified that a "radiation protection brief was required prior to accessing areas greater than seven feet." The workers were briefed on the RWPs by the radiation protection staff prior to commencing their activities; however, these two individuals failed to communicate to the radiation protection staff that they would be climbing on scaffolds above 7 feet. The first worker, a mechanic who was working on the feed water pump, climbed a scaffold across the pump area beyond the area that this individual was briefed and came across a shielded surge tank drain line, an area that radiation protection staff had not surveyed. The other worker, a contractor installing scaffold deck pans, climbed above seven feet from the area that he was briefed and came in contact with a Rad-Waste Max-recycle line, an area that also was not surveyed during the early stages of the Dresden Unit 3 outage. The mechanic came within 3 feet of a shielded drain line that read 100 mrem/hr on contact and this worker received an unplanned dose rate alarm of 41 mrem/hr, with a 40 mrem/hr dose rate alarm setting. The other individual, a contractor, climbed 7 feet on a scaffold above an area the he was briefed and received an unplanned dose rate alarm of 150 mrem/hr on contact with Rad Waste Max-recycle line. Follow-up surveys of the Max-recycle line indicated 156 mrem/hr on contact. The mechanic had a cumulative dose of 1 mrem and the contractor received 15 mrem from these entries.

Analysis:

The inspectors determined that the radiation workers' failure to comply with the requirements stated in the RWP were within the licensee's ability to foresee and correct 20 and should have been prevented, therefore constituted a performance deficiency. The performance deficiency was determined to be more-than-minor in accordance with Inspection Manual Chapter 0612, Appendix B, "Issue Screening," because the performance deficiency impacted 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.

Specifically, worker entry into areas beyond the RWP briefing could lead to unintended dose. The finding was determined to be of very-low safety significance (Green) in accordance with Inspection Manual Chapter 0609, Appendi x C, "Occupational Radiation Safety Significance Determination Process," dated August 19, 2008, because:

(1) it did not involve as-low-as-reasonably-achievable (ALARA) planning or work controls,

(2) there was no overexposure,

(3) there was no substantial potential for an overexposure, and

(4) the ability to assess dose was not compromised. The inspectors concluded that the cause of the finding involved a cross-cutting component in the human performance area of challenging the unknown because these individuals did not stop when faced with uncertain conditions. Risks were not evaluated and managed before proceeding. Specifically, a mechanic and a contract worker proceeded into areas that they were not briefed to enter which contained unknown dose rates. [H.11]

Enforcement:

Technical Specification 5.4.1, Procedures, states in part that "written procedures shall be established, implemented and maintained covering activities contained in Regulatory Guide 1.33, Revision 2, Appendix-A, dated February 1978."

NRC Regulatory Guide 1.33, Appendix A, Section 7 addresses "Procedures for Control Radioactivity" and Section 7e "Radiation Protection Procedures, section

(1) addresses Access Control to Radiation Areas Including Radiation Work Permits". RWPs DR-03-16-00816 and DR-03-16-00411 both specified that a "radiation protection brief was required prior to accessing areas greater than seven feet." Contrary to the above, on October 31, 2016, the licensee failed to obtain the required radiation protection brief prior to accessing areas greater than seven feet per RWPs DR-03-16-00816 and DR-03-16-00411.

Specifically, two technicians working in a radiation area under RWPs DR-03-16-00816 and DR-03-16-00411 entered into areas above 7 feet that were not discussed during RWP briefings. This caused the workers to receive unplanned dose rate alarms. Upon receiving the dose rate alarms, these individuals exited the area and immediately reported to the radiation protection staff. Because this violation was of very-low safety significance and was entered into the licensee's Corrective Action Program as IR 02735594 and IR 02735651 this violation is being treated as an NCV, consistent with Section 2.3.2 of the

Enforcement Policy. (NCV 05000237/2016004-01; 05000249/2016004-01: Failure to Comply With Radiation Work Permit Requirements Resulting In Unplanned Dose Rate Alarms)

.4 Problem Identification and Resolution (02.08)

a. Inspection Scope

The inspectors assessed whether problems associated with radiological hazard assessment and exposure controls were being identified at an appropriate threshold and were properly addressed for resolution. For select problems, the inspectors assessed the appropriateness of the corrective actions. The inspectors also assessed the licensee's program for reviewing and incorporating operating experience. The inspectors reviewed select problems related to human performance errors and assessed whether there was a similar cause and whether corrective actions taken

resolve the problems. The inspectors reviewed select problems related to radiation protection technician error and assessed whether there was a similar cause and whether corrective actions taken resolve the problems. These inspection activities constituted one complete sample as defined in IP 71124.01-05.

b. Findings

No findings were identified.

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

.1 Implementation of As-Low-As-Reasonably-Achievable and Radiological Work Controls

(02.04)

a. Inspection Scope

The inspectors reviewed the radiological administrative, operational, and engineering controls planned for selected radiologically significant work activities and evaluated the integration of these controls and as-low-as-reasonably-achievable (ALARA) requirements into work packages, work procedures and/or RWPs. The inspectors conducted observations of in-plant work activities and assessed whether the licensee had effectively integrated the planned administrative, operational, and engineering controls into the actual field work to maintain occupational exposure ALARA. The inspectors observed pre-job briefings, and determined if the planned controls were discussed with workers. The inspectors evaluated the placement and use of shielding, contamination controls, airborne controls, RWP controls, and other engineering work controls against the ALARA plans. The inspectors assessed licensee activities associated with work-in-progress to ensure the licensee was tracking doses, performed timely in-progress reviews, and, when jobs did not trend as expected, appropriately communicated additional methods to be used to reduce dose. The inspectors evaluated whether health physics and ALARA staff were involved with the management of radiological work control when in-field activities deviated from the planned controls. The inspectors assessed whether the Outage Control Center and station management provided sufficient support for ALARA re-planning.

22 The inspectors assessed the involvement of ALARA staff with emergent work activities during maintenance and when possible, attended in-progress review discussions, outage status meetings, and/or ALARA committee meetings. The inspectors compared the radiological results achieved with the intended radiological outcomes and verified that the licensee captured lessons learned for use in the next

outage. These inspection activities constituted one complete sample as defined in IP 71124.02-05.

b. Findings

No findings were identified.

.2 Radiation Worker Performance (02.05)

a. Inspection Scope

The inspectors observed radiation worker and radiation protection technician performance during work activities being performed in radiation areas, airborne radioactivity areas, or high radiation areas to assess whether workers demonstrated the ALARA philosophy in practice and followed procedures. The inspectors observed radiation worker performance to evaluate whether the training and skill level was sufficient with respect to the radiological hazards and the work involved. The inspectors interviewed individuals from selected work groups to assess their knowledge and awareness of planned and/or implemented radiological and ALARA work controls. These inspection activities constituted one complete sample as defined in IP 71124.02-05.

b. Findings

No findings were identified.

OTHER ACTIVITIES

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

4OA1 Performance Indicator Verification

.1 Safety System Functional Failures

a. Inspection Scope

The inspectors sampled licensee submittals for the Safety System Functional Failures (MS05) performance indicator for Dresden Nuclear Power Station, Units 2 and 3, for the period from the fourth quarter of 2015 through the third quarter of 2016. To determine the accuracy of the PI data reported during those periods, PI definitions and guidance 23 contained in the NEI Document 99-02, "Regulatory Assessment Performance Indicator Guideline," Revision 7, dated August 31, 2013, were used. The inspectors reviewed the licensee's operator narrative logs, MSPI derivation reports, issue reports, event reports and NRC Integrated Inspection Reports for the period of October 1, 2015, through September 30, 2016, to validate the accuracy of the submittals. The inspectors also reviewed the licensee's 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 Reactor Coolant System Leakage

a. Inspection Scope

The inspectors sampled licensee submittals for the RCS Leakage (BI02) performance indicator for Dresden Nuclear Power Station, Units 2 and 3, for the period from the fourth quarter of 2015 through the third quarter of 2016. To determine the accuracy of the PI data reported during those periods, PI definitions and guidance contained in the NEI

Document 99-02, "Regulatory Assessment Performance Indicator Guideline," Revision 7, dated August 31, 2013, were used. The inspectors reviewed the licensee's operator narrative logs, MSPI derivation reports, issue reports, event reports and NRC Integrated Inspection Reports for the period of October 1, 2015, through September 30, 2016, to validate the accuracy of the submittals. The inspectors also reviewed the licensee's 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 reactor coolant system leakage samples as defined in IP 71151-05.

b. Findings

No findings were identified.

.3 Occupational Exposure Control Effectiveness

a. Inspection Scope

The inspectors sampled licensee submittals for the Occupational Exposure Control Effectiveness (OR01) performance indicator for the period from the first quarter 2015 through the second quarter 2016. The inspectors used PI definitions and guidance contained in the NEI Document 99-02, "Regulatory Assessment Performance Indicator Guideline," Revision 7, dated August 2013, to determine the accuracy of the PI data reported during those periods. The inspectors reviewed the licensee's assessment of the PI for occupational radiation safety to determine if the indicator related data was adequately assessed and reported. To assess the adequacy of the licensee's PI data 24 collection and analyses, the inspectors discussed with radiation protection staff the scope and breadth of its data review and the results of those reviews. The inspectors independently reviewed electronic personal dosimetry dose rate and accumulated dose alarms and dose reports and the dose assignments for any intakes that occurred during the time period reviewed to determine if there were potentially unrecognized occurrences. The inspectors also conducted walkdowns of numerous locked high and very high radiation area entrances to determine the adequacy of the controls in place for these areas. Documents reviewed are listed in the Attachment to this report. This inspection constituted one occupational exposure control effectiveness sample as defined in IP 71151-05.

b. Findings

No findings were identified.

.4 Radiological Effluent Technical Specification/Offsite Dose Calculation Manual

Radiological Effluent Occurrences

a. Inspection Scope

The inspectors sampled licensee submittals for the Radiological Effluent Technical Specification/Offsite Dose Calculation Manual radiological effluent occurrences (PR01)performance indicator for the period from the first quarter 2015 through the second quarter 2016. The inspectors used PI definitions and guidance contained in the NEI Document 99-02, "Regulatory Assessment Performance Indicator Guideline," Revision 7, dated August 2013, to determine the accuracy of the PI data reported during those periods. The inspectors reviewed the licensee's issue report database and selected individual reports generated since this indicator was last reviewed to identify any potential occurrences such as unmonitored, uncontrolled, or improperly calculated effluent releases that may have impacted offsite dose. The inspectors reviewed gaseous effluent summary data and the results of associated offsite dose calculations for selected dates to determine if indicator results were accurately reported. The inspectors also reviewed the licensee's methods for quantifying gaseous and liquid effluents and determining effluent dose. Documents reviewed are listed in the Attachment to this report. This inspection constituted one Radiological Effluent Technical Specification/Offsite Dose Calculation Manual radiological effluent occurrences sample as defined in

IP 71151-05.

b. Findings

No findings were identified.

4OA2 Identification and Resolution of Problems

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

a. Inspection Scope

As 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 licensee's corrective action program at an appropriate threshold, adequate attention was being given to timely corrective actions, and adverse trends were identified and addressed. Some minor issues were entered into the licensee's corrective action program as a result of the inspectors' observations; however, they are not discussed in 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.

b. Findings

No findings were identified.

.2 Semi-Annual Trend Review

a. Inspection Scope

The inspectors performed a review of the licensee's corrective action program 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 corrective action program item screening discussed in Section 4OA2.1 above, licensee trending efforts, and licensee human performance results. The inspectors' review nominally considered the 6-month period of June 2016 through December 2016

, although some examples expanded beyond those dates where the scope of the trend warranted. The review also included issues documented outside the corrective action program 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 licensee's corrective action program trending reports. Corrective actions associated with a sample of the issues identified in the licensee's trending reports were reviewed for adequacy. b. Observations During the period covered in this inspection sample, the inspectors and the licensee noted similar trends in equipment and program performance. Although these errors did not always result in any immediate adverse consequences, a potential trend in these areas is apparent and suggests that additional licensee attention to affect corrective

26 actions may be appropriate. The licensee entered the following potential adverse trends into their CAP during this time period: steam leaks on Unit 2 (IR 3956785) and interlock door performance (IR 3955902). The Inspectors also identified steam leaks on Unit 3 as a potential adverse trend. Specific examples associated with these trends included, but were not limited to:

• On November 21, 2016, the U2 nuclear station operator was monitoring the 2B moisture separator drain tank (MSDT) leak repair via camera and noticed steam near the repair. The field supervisor verified the steam was coming from the Furmanite temporary repair of the 2B MSDT pipe plug. The original leak was identified on May 27, 2016, (IR 2674900) and a temporary repair (Furmanite box)was installed, which started leaking again on June 15, 2016, (IR 2681796).

Additionally, there was a steam leak on the 2A MSDT (IR 2736509) which the licensee took action on December 10, 2016, to repair both leaks with a planned down power to less than 20 percent.

• On November 8, 2016, during Unit 3 Division 1 undervoltage (UV) testing, secondary containment differential pressure decreased to less than -0.25 in Water Column vacuum when SBGT started and reactor building ventilation secured. This resulted in a loss of secondary containment and entry into Technical Specification LCO 3.6.4.1 Condition A for Unit 2 which was in Mode 1 (IR 2738559). The loss of secondary containment vacuum required the licensee to make an Emergency Notification System report (52355) for an event or condition that could have prevented the fulfillment of a safety function. During the UV testing the inner door of the Unit 2 reactor building material interlock was open in support of D3R24 which meant the exterior door was the secondary

containment boundary. The exterior door was degraded resulting in excessive leakage. It was recently noted that the seals were badly damaged on the exterior door, but no IR was written to document the degraded condition. The interior reactor building material interlock door was closed and reactor building differential pressure was restored below TS limits. The licensee subsequently replaced the outer door seals.

• On October 14, 2016, operators in the plant noted a loud noise coming from the Unit 3 high pressure heater bay. Subsequent investigation identified a steam leak on a line from the high pressure turbine to the 3D moisture separator. The steam plume was directed straight up about 8-10 feet toward the ceiling. The steam was not directly impinging on any plant components or electrical equipment and no water was accumulating in the area which could adversely affect adjacent equipment. The condition was monitored and repaired in November 2016 during the D3R24 outage. This review constituted one semi-annual trend review inspection sample as defined in IP 71152.

c. Findings

No findings were identified.

.3 Annual Follow-up of Selected Issues:

Review of Final Corrective Actions Associated with Design Control Violation, Non-Cited Violation 05000237/2016001-01; 05000249/2016001-01, (Failure to Maintain Design Control of the 2/3 Emergency Diesel Generator)

a. Inspection Scope

The inspectors performed a review of the licensee's CAP and associated documents, specifically IR 2593932, "2/3 EDG Ventilation System Dampers Do Not Failsafe Open." The inspectors interviewed personnel, performed walkdowns, reviewed design change documents, observed the installation of plant modifications, and verified the completion of and assessed the adequacy of plant design corrective actions taken in response to a loss of control pneumatics to the 2/3 EDG room ventilation system on November 6, 2015. The inspectors' review and evaluation was focused on the licensee's corrective actions to ensure they: were complete, accurate, and timely; considered extent of condition; provided appropriate classification and prioritization; provided identification of root and contributing causes; were appropriately focused; included action taken which resulted in the correction of the identified problem; identified negative trends; ensured operating experience was adequately evaluated for applicability; and communicated applicable lessons learned to appropriate organizations. The inspectors noted that the licensee's corrective actions addressed deficiencies in the design of the 2/3 EDG ventilation system, specifically creating fail-safe system response for the inlet and exhaust dampers to a loss of non-safety related pneumatic supply air and back-up nitrogen. This review constituted one follow-up inspection sample for in-depth review as defined in IP 71152-05. b. Background On November 6, 2015, with the normal source of pneumatics to the 2/3 EDG room ventilation damper positioning system secured for maintenance, on two occasions for approximately twenty minutes each the back-up non-safety related nitrogen source depressurized causing the room ventilation dampers to fail in a closed condition. Based on the licensee's understanding of system performance a loss of 2/3 EDG room ventilation damper control pneumatics was supposed to result in the dampers failing conservatively open. The licensee had not previously tested the performance of the back-up nitrogen system nor had they tested EDG room ventilation system response to a complete loss of pneumatics to ascertain actual system response. The loss of 2/3 EDG ventilation and therefore the inability of the 2/3 EDG to be able to complete its operational mission time during a design basis accident due to a loss of non-safety related pneumatic control air was originally documented as a NCV of 10 CFR Part 50, Appendix B, Criterion III, "Design Control," in NRC Integrated Inspection Report 05000237/2016001; 05000249/2016001 (ADAMS Ascension Number ML16120A618).

28 c. Observations As discussed in the "Inspection Scope" section above, the inspectors' review was focused on the licensee's design and installation of a plant design modification to create a fail-safe configuration for the 2/3 EDG ventilation system during a loss of operating pneumatics to the inlet and exhaust ventilation dampers. The inspectors noted that the design modification performed by the licensee adequately addressed the cause of the event which resulted in a NCV of 10 CFR Part 50, Appendix B, Criterion III, "Design Control," and would enable the 2/3 EDG to achieve its safety mission during a design basis accident. The inspectors review of this plant modification included an assessment of whether the 2/3 EDG ventilation retained its automatic isolation capability during fire conditions in the 2/3 EDG room. The inspectors noted that the modification retained the requirement that the inlet and exhaust dampers close upon a fire detection system signal in the 2/3 EDG room in accordance with the site's Fire Protection Report and National Fire Protection Association (NFPA) codes and standards (NFPA 12: Standard on Carbon Dioxide Extinguishing Systems).

d. Findings

No findings were identified.

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

.1 (Closed) Supplemental Licensee Event Report 05000237/2016-002-01, "Unit 2 HPCI Inlet Steam Drain Pot Piping Leak Resulting in HPCI Inoperability"

a. Inspection Scope

The inspectors reviewed the licensee's supplemental response to and assessment of a through-wall leak which developed on the Unit 2 HPCI inlet drain pot drain piping. Specifically, on May 16, 2016, while in standby operation, a through-wall steam leak was observed coming from the Unit 2 HPCI 1 inch diameter inlet drain pot drain piping upstream of the main condenser return isolation valve 2-2301-29. The leak was identified to be from 2-2323-1", which is ASME Code Class 2 piping. Due to the piping being ASME Code Class 2, it was required to be isolated in accordance with Technical Requirements Manual 3.4.a, Structural Integrity. Isolating this piping resulted in the Unit 2 HPCI system becoming inoperable. Follow-up investigation and testing of the failed component indicated a failure mechanism of liquid droplet impingement. The piping that failed was believed to have been replaced in 2013 with stainless steel, a material resistant to liquid drop impingement, but was not replaced due to a work package revision/scope removal change prior to execution. Additional information provided by this Supplemental LER included details on the licensee inadvertently removing the affected pipe replacement from the work schedule and extent of condition reviews for work scope revisions.

The inspectors review of the licensee's initial LER for this issue was documented in NRC Integrated Inspection Report 05000237/2016003; 05000249/2016003 (ADAMS ascension number ML16298A205) and resulted in a self-revealed finding of very low safety significance (Green) and associated NCV of TS 5.4.1.a, "Procedures."

29 The licensee reported this event in accordance with 10 CFR 50.73(a)(2)(v)(D), any event or condition that could have prevented the fulfillment of the safety function of structures or systems that are needed to mitigate the consequences of an accident. Documents reviewed are listed in the Attachment to this report. This LER is closed. This event follow up review constituted one sample(s) as defined in IP 71153-05.

b. Findings

No findings were identified.

4OA5 Other Activities

.1 (Closed) Apparent Violation 05000237/2013407-01; 05000249/2013407-01:

Failure to Satisfy Access Authorization Program Requirements Involving Licensed Operators On October 28, 2013, the U.S. Nuclear Regulatory Commission (NRC) issued Confirmatory Order EA-13-068 (Order) (ADAMS ascension number ML13298A144) in lieu of enforcement action to the licensee. As a result of a National Labor Relations Board hearing and settlement, on May 4, 2015, the Order was relaxed to allow Exelon additional time to complete the required actions in V.A.1, V.A.2, and V.A.3 (ADAMS ascension number ML15125A103). The elements of the agreement between the NRC and the licensee consisted of the following five specific actions: Item V.A.1: By November 30, 2015, revise Exelon procedure SY-AA-103-513, "Behavioral Observation Program":

(1) to provide additional guidance on the types of offsite activities, if observed, or credible information that should be reported to reviewing officials; and

(2) to ensure that procedural requirements to pass information forward without delay are clearly communicated. Item V.A.2: By January 15, 2016, Exelon shall provide training to Exelon staff on the revision described in V.A.1. Item V.A.3: By May 31, 2016, Exelon shall develop and conduct an effectiveness assessment of its revised procedure and training to determine if Exelon personnel remain aware of the need to report observed offsite aberrant behavior or credible information. Item V.B: Within 90 days of the effective date of the Confirmatory Order, Exelon will develop and make a presentation based on the facts and lessons learned from the events that gave rise to the Confirmatory Order. Exelon agrees to make this presentation at an appropriate industry forum and to submit an operating experience summary to an industry-wide organization.

Exelon will make the presentation materials available to the onsite NRC resident inspectors at the Dresden Station. Item V.C: Unless otherwise specified, Exelon will submit a written status of the Confirmatory Order action items to the NRC Region III Director of Reactor Safety, by October 31, 2014, and annually thereafter, until all actions are completed.

30 The inspectors independently verified that the required actions listed above were completed. Specifically, the inspectors reviewed licensee records and conducted interviews with both plant management and selected plant staff to verify that:

• On November 18, 2015, licensee procedure SY-AA-103-513, "Behavioral Observation Program," Revision 12 was issued. This revision contained additional guidance on reporting and forwarding credible information.

(Item V.A.1.)

• On January 14, 2016, a supervisory training brief was given to all staff regarding the changes to procedure. Exelon-specific training regarding the Behavioral Observation Program (BOP) was added to the Generic Plant Access Training to ensure that the BOP training will continue. (Item V.A.2.)

• On May 31, 2016, Exelon Corporate staff completed an effectiveness review of SY-AA-103-513 and concluded that the procedure was effective. The inspectors also discussed the procedure changes with selected plant staff and independently concluded that the procedure was effective. (Item V.A.3)

• On December 4, 2013, the licensee prepared and made a presentation at the Nuclear Energy Institute Advisory Task Force Meeting. The Resident Inspectors were provided a copy of the briefing materials. (Item V.B.)

• Exelon provided written status updates to the NRC Region III Director of Reactor Safety on September 30, 2014, October 28, 2015, and October 13, 2016. (Item V.C.) Based on the licensee's actions described above, and in accordance with Confirmatory Order EA-13-068 as revised on May 4, 2015, the NRC has completed its review of the licensee's implementation of the conditions of the Order. In addition, Apparent Violation 05000237/2013407-01; 05000249/2013407-01 is closed.

4OA6 Management Meetings

.1 Exit Meeting Summary On January 13, 2017, the inspectors presented the inspection results to Mr. J. Washko, and other members of the licensee staff.

The licensee acknowledged the issues presented. The inspectors confirmed that none of the potential report input discussed was considered proprietary.

.2 Interim Exit Meetings Interim exits were conducted for:

• The inspection results for the Radiation Safety Program review with Mr. P. Karaba, Site Vice President, on November 4, 2016.

• The results of the Inservice inspection with Station Plant Manager, Mr. J. Washko, on November 8, 2016.

• The results of the Emergency Preparedness inspection with Regulatory Assurance Manager, Mr. B. Franzen, on December 20, 2016.

31

• The results of the Emergency Preparedness Program inspection with Mr. D. Doggett, Emergency Preparedness Manager, conducted over the phone on December 20, 2016. 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.

.3 Management Briefing The security inspector presented the results of a review regarding the licensee's activities in response to a Confirmatory Order.

Mr. B. Franzen, and other members of licensee management attended the briefing on November 29, 2016. The inspector asked the licensee whether any materials examined during the inspection should be considered proprietary. No proprietary information was identified.

4OA7 Licensee-Identified Violations The following violation of very low significance (Green) was identified by the licensee and is a violation of NRC requirements which meets the criteria of the NRC Enforcement

Policy for being dispositioned as a Non-Cited Violation (NCV).

• Title 10 of the Code of Federal Regulations (10 CFR) 50.54(q)(2) requires that a holder of a nuclear power reactor operating license follow and maintain the effectiveness of an emergency plan that meets the requirements in 10 CFR Part 50, Appendix E and the planning standards of 10 CFR 50.47(b).

Title 10 CFR Part 50.47(b)(4) states, "A standard emergency classification

and action level scheme, the bases of which include facility system and effluent parameters, is in use by the nuclear facility licensee, and State and local response plans call for reliance on information provided by facility licensees for determinations of minimum initial offsite response measures." Contrary to the above, between April 2013, and February 2016, the licensee failed to maintain the effectiveness of the emergency plan by failing to maintain the effluent parameters contained in the standard emergency classification and action level scheme. Specifically, the standard emergency classification and action level scheme associated with the radiological effluents at Dresden Nuclear Power Station was not updated to reflect the changes in the X/Q dispersion factor that were made during the April 2013, Offsite Dose Calculation Manual revision. Consequently, the effluent monitor emergency classification and action level thresholds were non-conservative by a factor of 3.8 until this condition was identified and corrected by Dresden Nuclear Power Station in February 2016. The inspectors determined that the finding was of very low significance (Green) in accordance with NRC Inspection Manual Chapter 0609, Appendix B, "Emergency Preparedness Significance Determination Process, Figure 5.4-1, because the emergency action level classification of an Unusual Event, RU1, would be declared in a degraded manner, not within the required 15 minutes.

32 The emergency action level classification for the Alert, Site Area Emergency, and General Emergency (RA1, RS1, and RG1) would still be capable of being declared in timely manner, within 15 minutes, using alternate conditions within the emergency action level. Because this finding is of very low safety significance, and has been entered into Exelon's CAP under IR 02652711, this violation is being treated as a Green NCV consistent with Section 2.3.2 of the NRC's Enforcement Policy. ATTACHMENT:

SUPPLEMENTAL INFORMATION

KEY POINTS OF CONTACT

Licensee

P. Karaba, Site Vice President

J. Washko, Station Plant Manager

D. Anthony, NDE Services Manager

L. Antos, Manager Site Security

C. Bachman, Plant Engineering

R. Bauman, Shift Operations Superintendent

M. Budelier, Senior Engineering Manager

H. Bush, Development Manager

J. Condreay, Operations Training Instructor

T. Dean, Director, Site Training

D. Doggett, Emergency Preparedness Manager

B. Franzen, Regulatory Assurance Manager

F. Gogliotti, Director, Site Engineering

P. Hansett, Work Control Director

R. Johnson, Chemistry

D. Ketchledge, Engineering

J. Kish, Site ISI Lead

K. Kretsinger, Security Operations Supervisor

S. Matzke, Corrective Action Program Coordinator

A. McMartin, Manager Site Chemistry, Environment & Radwaste

J. Miller, NDE Level III

G. Morrow, Operations Director

M. Overstreet, Radiation Protection Manager

M Pavey, Senior Health Physicist

T. Pille, Security Training Supervisor

J. Quinn, Director, Site Maintenance

W. Remiasz, Outage Manager

B. Sampson, OR Manager

D. Thomas, Training Manager

D. Walker, Regulatory Assurance - NRC Coordinator

U.S. Nuclear Regulatory Commission J. Cameron, Chief, Reactor Projects, Branch 4

IEMA

M. Porfirio, Resident Inspector, Illinois Emergency Management Agency

L. Torres, ASME Inspector, Illinois Emergency Management Agency

LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED

Opened

05000237/2016004-01

05000249/2016004-01 NCV Failure to Comply With Radiation Work Permit Requirements Resulting In Unplanned Dose Rate Alarms

(Section 2RS1.3)

Closed

05000237/2016004-01

05000249/2016004-01 NCV Failure to Comply With Radiation Work Permit Requirements Resulting In Unplanned Dose Rate Alarms (Section 2RS1.3)

05000237/2013407-01

05000249/2013407-01 AV Failure to Satisfy Access Authorization Program Requirements Involving Licensed Operators

(Section 4OA5.1)

05000237/2016-002-01 LER Unit 2 HPCI Inlet Steam Drain Pot Piping Leak Resulting

in HPCI Inoperability (Section 4OA3.1)

Discussed

05000237/2016001-01

05000249/2016001-01 NCV Failure to Maintain Design Control of the 2/3 Emergency Diesel Generator (Section 4OA2.3)

LIST OF DOCUMENTS REVIEWED

The following is a partial list of documents reviewed during the inspection.

Inclusion on this list does not imply that the NRC inspector reviewed the documents in their entirety, but rather that