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Nine Mile Point Nuclear Station - Integrated Inspection Report 05000220/2016002 and 05000410/2016002 and Exercise of Enforcement Discretion
	ML16221A101
Person / Time
Site:	Nine Mile Point
Issue date:	08/08/2016
From:	Scott M L; Division Reactor Projects I
To:	Bryan Hanson; Exelon Generation Co, Exelon Nuclear
References
	EA-16-111 05000220/2016002 and 05000410/2016002, IR 2016002
	Download: ML16221A101 (53)
	v • d • e

August 8, 2016

EA-16-111 Mr. Bryan Senior Vice President, Exelon Generation Company, LLC President and Chief Nuclear Officer, Exelon Nuclear 4300 Winfield Road Warrenville, IL 60555

SUBJECT: NINE MILE POINT NUCLEAR STATION - INTEGRATED INSPECTION REPORT 05000220/2016002 AND 05000410/2016002 AND EXERCISE OF ENFORCEMENT DISCRETION

Dear Mr. Hanson:

On June 30, 2016, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at the Nine Mile Point Nuclear Station, LLC (NMPNS), Units 1 and 2. The enclosed inspection report documents the inspection results, which were discussed on July 13, 2016, with Mr. Peter Orphanos, Site Vice President, and other members of the NMPNS staff.

NRC inspectors examined activities conducted under your licenses as they relate to safety and compliance with the Commission's rules and regulations and with the conditions of your licenses. The inspectors reviewed selected procedures and records, observed activities, and interviewed personnel.

The inspectors documented three findings of very low safety significance (Green) in this report.

Two of these findings involved a violation of NRC requirements. The NRC is treating these violations as non-cited violations (NCVs) consistent with Section 2.3.2.a of the Enforcement Policy. In addition, one licensee-identified issue of low to moderate safety significance (White) was determined to be a violation of NRC requirements. The NRC determined this issue warranted enforcement discretion because it satisfied the criteria established in NRC Enforcement Policy Section 9.1, "Enforcement Discretion for Certain Fire Protection Issues (10 CFR 50.48)," and NRC Inspection Manual Chapter 0305, "Operating Reactor Assessment Program," Section 11.05(b), "Treatment of Items Associated with Enforcement Discretion." The Regional Administrator, Region I, was consulted regarding enforcement discretion for this issue.

If you contest the NCVs in this report, you should provide a response within 30 days of the date of this inspection report, with the basis for your denial, to the U.S. Nuclear Regulatory Commission, ATTN: Document Control Desk, Washington, DC 20555-0001; with copies to the Regional Administrator, Region I; the Director, Office of Enforcement, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001; and the NRC Resident Inspectors at NMPNS. In addition, if you disagree with the cross-cutting aspect assigned to any finding, or a finding not associated with a regulatory requirement 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 I, and the NRC Resident Inspectors at NMPNS. In accordance with Title 10 of the Code of Federal Regulations 2.390 of the NRCs "Rules of Practice," a copy of this letter, its enclosure, and your response (if any) will be available electronically for public inspection in the NRC's Public Document Room or from the Publicly Available Records component of the NRC's Agencywide Documents Access 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 Scott, Director Division of Reactor Projects Docket Nos. 50-220 and 50-410 License Nos. DPR-63 and NPF-69

Enclosure:

Inspection Report 05000220/2016002 and 05000410/2016002

w/Attachment:

Supplementary Information

cc w/encl: Distribution via ListServ

SUMMARY

Inspection Report 05000220/2016002 and 05000410/2016002; 04/01/2016 - 06/30/2016; NMPNS, Units 1 and 2; Adverse Weather Protection, Operability Determinations and Functionality Assessments, and Radiological Hazard Assessment and Exposure Controls.

This report covered a 3-month period of inspection by resident inspectors and announced inspections performed by regional inspectors. The inspectors identified three Green findings.

Two of these findings also involved violations of NRC requirements and are being treated as non-cited violations (NCVs). The significance of most 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 U.S. Nuclear Regulatory Commission (NRC) requirements are dispositioned in accordance with the NRC's Enforcement Policy, dated February 4, 2015. The NRC's program for overseeing the safe operation of commercial nuclear power reactors is described in NUREG-1649, "Reactor Oversight Process," Revision 5.

Cornerstone: Initiating Events

Green.

The inspectors identified a Green finding (FIN) of PI-AA-125, "Corrective Action Program," Revision 3, when Exelon failed to implement adequate corrective actions in March 2003, to prevent water intrusion into the Unit 2 normal switchgear building area at elevation 237. Specifically, on May 4, 2016, the inspectors observed water leaking into the normal switchgear room through a wall on elevation 237. The leakage was through a section of the wall that contained electrical junction boxes that were not sealed. The water progressed under inverter 2BYS-SWG001B, which led to the potential for a reactor scram from an electrical fault associated with uninterruptible power supply battery breakers. Previously, a reactor scram had occurred at Unit 2 on March 4, 2014, when the inverter was lost because of an electrical fault, as such this was a known initiating event single point vulnerability . Corrective actions included entering the issue into the corrective action program (CAP) (IR 02664534), generating work order (WO) C93414574 to seal or repair the wall, and installing temporary barriers to redirect any water away from the switchboard. The WO is scheduled to be performed in October 2016 with an action to assess moving the work to the refueling outage if needed to remove the electrical junction boxes to apply coating to the wall. The finding is more than minor because it is associated with the Protection Against External Factors attribute of the Initiating Events cornerstone and affected the cornerstone objective of limiting the likelihood of those events that upset plant stability and challenge critical safety functions during power operations. Specifically, Exelon did not ensure the surface area behind the electrical junction boxes was coated to prevent water intrusion into the normal switchgear room at elevation 237. The water intrusion through this area of the wall had the potential to cause an electrical fault on 2BYS-SWG001B resulting in a reactor scram similar to the reactor scram in March 2014. The inspectors evaluated this finding using IMC 0609.04, "Initial Characterization of Findings," and Exhibit 1 of IMC 0609, Appendix A, "The Significance Determination Process for Findings At-Power." The inspectors determined that this finding was of very low safety significance (Green) because it did not represent the potential for both a reactor scram and a loss of mitigation equipment relied upon to transition the plant from the onset of the trip to a stable shutdown condition. The inspectors did not assign a cross-cutting aspect to this finding because the performance deficiency occurred greater than three years ago; therefore, it is not considered to be indicative of current plant performance. (Section 1R01)

Cornerstone: Mitigating Systems

Green.

The inspectors identified a Green NCV of Unit 2 Technical Specification (TS) 3.5.1, "Emergency Core Cooling (ECCS) Systems-Operating," and TS 3.5.3, "Reactor Core Isolation Cooling (RCIC) System," for failure to ensure all necessary attendant instrumentation required for the systems to perform their specified safety functions were capable of performing their related support function in all require modes of applicability. Specifically, the inspectors identified the Unit 2 wide range level indication to be inaccurate during Mode 2 and at 200 pounds per square inch gauge (psig) reactor pressure, a mode of applicability requiring both high-pressure core spray (HPCS) and RCIC to be operable. This resulted in a high level trip signal being locked preventing HPCS or RCIC from auto initiating, rendering the systems inoperable. Upon identification, Exelon generated issue report (IR) 02667837 to address the inspectors' concern regarding the wide range level indication. An action was created to evaluate the impact of the wide range level discrepancy with regard to its impact on safety-related functions to supply water in the TS Mode of Applicability. Exelon also plans to assess the need for a TS amendment. The performance deficiency was more than minor because it was associated with the Equipment Performance attribute of the Mitigating Systems cornerstone and affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, Exelon failed to recognize that the wide range level indication did not provide accurate indication at low reactor pressures and temperatures, preventing automatic safety-related functions associated with high drywell pressure automatic initiation signals and manual start functions.

This would require operators to manually open the HPCS and RCIC injection valves during these conditions should a loss of offsite power or loss-of-coolant accident occur. In accordance with IMC 0609.04, "Initial Characterization of Findings," and Exhibit 2 of IMC 0609, Appendix A, "The Significance Determination Process for Findings At-Power," the inspectors determined that the finding was of very low safety significance (Green), because the finding was not a design or qualification deficiency, did not represent a loss of system safety function, and did not screen as potentially risk significant due to seismic, flooding, or severe weather initiating event. The finding had a cross-cutting aspect in the area of Problem Identification and Resolution, Identification. Exelon personnel had many opportunities, including during the reactor startup in May of 2016, to question operability of the instrumentation that provides input for automatic initiation and isolation signals. As a result, Exelon personnel failed to identify that the wide range level indication did not support operability of the HPCS and RCIC systems during reactor startup on May 5, 2016. [P.1] (Section 1R15)5

Cornerstone: Occupational Radiation Safety

Green.

A self-revealing NCV of TS 5.4.1 "Procedures" was identified when a worker performed a radiological work activity without notifying radiation protection personnel and, as a result, did not comply with procedure RP-AA-1008, "Unescorted Access to and Conduct in Radiologically Controlled Areas, Revision 5," in being briefed on the necessary radiological work controls and conditions for performance of the Unit 2 reactor seal cleaning work activity. Specifically, on April 11, 2016, a worker entered the Unit 2 reactor cavity to perform inspection of the reactor seal that was highly contaminated. Although not previously discussed with radiation protection staff, the worker cleaned the highly contaminated reactor seal with rags and carried the highly contaminated rags (5 rem/hr) in his hand out of the reactor cavity, which resulted in unplanned radiation exposure to the worker's hand. Exelon's immediate corrective actions included reinforcing the need to properly communicate radiological work activities with radiation protection, and require workers to carry WOs with them to improve communications with radiation protection. Exelon entered the issue into the corrective action program (CAP) as IR 02654591. The failure of the worker to discuss the full scope of the radiological work activity with radiation protection staff, who were subsequently not effectively briefed on the expected radiological work conditions and requisite radiological controls needed for the work activity, is a performance deficiency that was reasonably within Exelon's ability to foresee and correct. The finding was determined to be more than minor because it affected the human performance attribute of the Occupational Radiation Safety cornerstone objective. Using IMC 0609, Appendix C, "Occupational Radiation Safety Significance Determination Process," the finding was determined to be of very low safety significance (Green) because it did not involve: (1) as low as reasonably achievable (ALARA) occupational collective exposure planning and controls, (2) an overexposure, (3) a substantial potential for overexposure, or (4) an impaired ability to assess dose. The finding is self-revealing because Exelon was made aware of the situation when an air monitor alarmed.

The finding had a cross-cutting aspect of Human Performance, Team Work, since individuals and work groups did not communicate and coordinate their activities within and across organizational boundaries to ensure nuclear safety was maintained. Specifically, the worker did not adequately communicate to radiation protection staff, the reactor seal cleaning activity to be performed. As a result, radiation protection personnel did not prescribe sufficient radiological controls for this high-contamination work activity, and led to an unintended exposure to the worker's hand. [H.4] (Section2RS1)6 Attachment

REPORT DETAILS

Summary of Plant Status Unit 1 began the inspection period at 100 percent power. On May 13, 2016, operators reduced reactor power to 85 percent to restore reactor recirculation motor generator set (RRMG) 14 following planned maintenance. Operators restored power to 100 percent on May 14. On May 21, operators reduced reactor power to 85 percent to restore RRMG set 15 following planned maintenance and to perform a control rod pattern adjustment. Operators restored power to 100 percent the same day. On June 12, operators reduced reactor power to 85 percent to restore RRMG set 15 following maintenance and to perform turbine valve testing.

Operators restored power to 100 percent the same day.

Unit 2 began the inspection period at 100 percent power. On April 10, 2016, operators began reducing power to remove Unit 2 from service in order to conduct a refueling and maintenance outage. Unit 2 was removed from the grid on April 11. On May 4, operators commenced a reactor startup. Unit 2 was synchronized to the grid on May 5, concluding the refuel and maintenance outage. Unit 2 reached 100 percent power on May 7. On May 28, operators reduced reactor power to 85 percent to perform turbine valve testing and to perform a rod pattern adjustment. Operators restored power to 100 percent the same day.

REACTOR SAFETY

Cornerstones: Initiating Events, Mitigating Systems, and Barrier Integrity

1R01 Adverse Weather Protection

Main article: IP 71111.01

.1 Readiness for Seasonal Extreme Weather Conditions

a. Inspection Scope

The inspectors performed a review of Exelon's readiness for the onset of seasonal high temperatures. The review focused on the Unit 2 diesel generator rooms. The inspectors reviewed the Updated Final Safety Analysis Report (UFSAR), TSs, control room logs, and the CAP to determine what temperatures or other seasonal weather could challenge these systems and to ensure Exelon personnel had adequately prepared for these challenges. The inspectors reviewed station procedures, including Exelon's seasonal weather preparation procedure and applicable operating procedures. The inspectors performed walkdowns of the selected systems to ensure station personnel identified issues that could challenge the operability of the systems during hot weather conditions.

Documents reviewed for each section of this inspection report are listed in the Attachment.

b. Findings

No findings were identified.

.2 Summer Readiness of Offsite and Alternate Alternating Current (AC) Power Systems

a. Inspection Scope

The inspectors reviewed Unit 1 and Unit 2 plant features and procedures for the operation and continued availability of the offsite and alternate AC power systems to evaluate readiness of the systems prior to seasonal high grid loading during the week of May 4, 2016. The inspectors reviewed Exelon's procedures affecting these areas and the communications protocols between the transmission system operator and Exelon implemented since the previous sample in 2015. This review focused on changes to the established program and material condition of the offsite and alternate AC power equipment. The inspectors assessed whether Exelon established and implemented appropriate procedures and protocols to monitor and maintain availability and reliability of both the offsite AC power system and the onsite alternate AC power system. The inspectors evaluated the material condition of the associated equipment by reviewing IRs and walking down portions of the offsite and AC power systems including the 345 kilovolt (kV) and 115 kV switchyards.

b. Findings

No findings were identified.

.3 External Flooding (2 samples)

a. Inspection Scope

On April 29, 2016, the inspectors performed an inspection of the external flood protection measures for NMPNS. The inspectors reviewed TSs, procedures, design documents, and the UFSAR/USAR, which depicted the design flood levels and protection areas containing safety-related equipment to identify areas that may be affected by external flooding. The inspectors conducted a general site walkdown of the site exterior flood berm during the week of April 25, 2016 and external water intrusion to Unit 2 normal switchgear room elevation 237 feet on May 3, 2016 to ensure that Exelon erected flood protection measures in accordance with design specifications. Where applicable the inspectors determined installed flood seal service life and verified that adequate procedures existed for inspecting the installed seals.

b. Findings

Introduction.

The inspectors identified a Green finding (FIN) of PI-AA-125, "Corrective Action Program," Revision 3, when Exelon failed to implement adequate corrective actions in March 2003, to prevent water intrusion into normal switchgear building elevation 237. Specifically, on May 4, 2016, the inspectors observed water leaking into the Unit 2 normal switchgear room through a wall on elevation 237. The leakage was through a section of the wall that contained electrical junction boxes which were not sealed. The water progressed under inverter 2BYS-SWG001B, which led to the potential for a reactor scram from an electrical fault associated with UPS battery breakers. This was a known initiating event single point vulnerability, since a reactor scram had previously occurred at Unit 2 on March 4, 2014, when the inverter was lost because of an electrical fault.

Description.

On May 3, 2016, Exelon generated IR 02664534, which identified water intrusion through a wall into the Unit 2 normal switchgear building at elevation 237. When the inspectors examined the area they noted inverter 2BYS-SWG001B was surrounded by water that had emanated through the north facing wall. Inverter 2BYS-SWG001B provides direct current (DC) power to uninterruptible power supplies that ensure a reliable power supply to critical plant loads. The inspectors also noted that 2VBB-UPS1D and 2VBB-UPS3B battery breakers were wetted at their lower surface from the water intrusion. Inverter 2BYS-SWG001B provides a DC source of power to a number of non-safety related UPSs. The UPSs utilize three sources of power: (1) a normal AC source, (2) a maintenance AC source, and (3) a DC source, to ensure a reliable source of power to its respective loads. Although battery switchgear 2BYS-SWG001B is non-safety related, it is a known initiating event single point vulnerability since an electrical fault on 2VBB-UPS3B can lead to a reactor scram. On March 4, 2014, an electrical fault on UPS3B occurred and resulted in loss of cooling to the reactor recirculation pumps and a subsequent manual reactor scram. This March 2014 scram was documented in IR 01700561.

Water intrusion into the 237 foot elevation of the normal switchgear building has been an ongoing concern that has not been adequately addressed. On March 9, 2000, NMPNS documented water intrusion into the area in IR 02091644 (DER 2-2000-0904) to evaluate and correct water intrusion near inverter 2BYS-SWG001B at normal switchgear building elevation 237. The evaluation that was performed under IR 02091644 determined that the appropriate corrective action would be to seal the entire wall with a coating. In March of 2003, WO 02-01974-00 was performed, which resulted in coating being applied to the wall.

The inspectors noted during their walkdown that a portion of the wall which contained electrical junction boxes was not sealed. Water was emanating from the unsealed area, and wetting the battery switchgear. Thus the original non-conforming condition was uncorrected. PI-AA-125, "Corrective Action Program," Section 4.3.6.3 states "If CAPRs/FWAPs or CAs/FWACs are determined to be ineffective, then a CR shall be originated to document the deficiency and initiate resolution." Contrary to this, the section behind the electrical junction boxes that was not painted continued to provide a path of water intrusion into the area allowing the possibility of an electrical fault and reactor scram.

The inspectors also identified that in April 2014, NMPNS documented in IR 02006312 that water was leaking through the wall in the same area.

As of June 30, 2016, the WO that was created to address the concern, C92661196, had not been scheduled to be performed, two years after the issue was documented.

The May 2016 IR 02664534 was screened to be addressed through a new WO, C93414574, scheduled to be performed in October 2016, with an action to assess moving the work to the refueling outage if needed to remove the electrical junction boxes to apply coating to the wall.

Analysis.

The inspectors determined that Exelon's failure, in March 2003, to complete corrective actions necessary to prevent water intrusion into elevation 237 of the normal switchgear building, a non-conforming condition, in accordance with PI-AA-125, "Corrective Action Program," was a performance deficiency that was reasonably within Exelon's ability to foresee and correct and should have been prevented. This finding is more than minor because it is associated with the Protection Against External Factors attribute of the Initiating Events cornerstone and affected the cornerstone objective of limiting the likelihood of those events that upset plant stability and challenge critical safety functions during power operations. Specifically, Exelon did not complete corrective actions to ensure the surface area behind the electrical junction boxes was coated to prevent water intrusion into the 237 foot elevation of the normal switchgear building. This non-conforming condition still exists and has the potential to cause an electrical fault on 2BYS-SWG001B, a known single point vulnerability which may cause a plant transient such as the reactor scram in March 2014.

The inspectors evaluated this finding using IMC 0609.04, "Initial Characterization of Findings," and Exhibit 1 of IMC 0609, Appendix A, "The Significance Determination Process for Findings At-Power," issued June 19, 2012. The inspectors determined that this finding was of very low safety significance (Green) because it did not have the potential to result in both a reactor scram and a loss of mitigation equipment relied upon to transition the plant from the onset of the trip to a stable shutdown condition.

The inspectors did not assign a cross-cutting aspect to this finding because the performance deficiency occurred greater than three years ago; therefore, it is not considered to be indicative of current plant performance.

Enforcement.

Enforcement action does not apply because this performance deficiency did not involve a violation of a regulatory requirement. Specifically, the structures, systems, and components (SSCs) in this event, the normal switchgear wall on elevation 237 and battery switchgear 2BYS-SWG001B, are not safety related. This issue was entered into Exelon's CAP as IR 02664534. Because this finding does not involve a violation and is of very low safety significance, it is identified as a finding. (FIN 05000410/2016002-01, Ineffective Corrective Action Results in Water Intrusion to Battery Switchgear Room) 1R04 Equipment Alignment

.1 Partial System Walkdowns

Main article: IP 71111.04

a. Inspection Scope

The inspectors performed partial walkdowns of the following systems:

Unit 2 'B' shutdown cooling system during operation to support cold shutdown on April 11, 2016 Unit 2, Division II, emergency diesel generator (EDG) and switchgear during Division I switchgear maintenance window on April 20, 2016 Unit 2, Division I, control building ventilation system during Division II control building chiller maintenance window on May 23, 2016 Unit 2 standby gas treatment system on June 7, 2016 Unit 1 liquid poison system on June 7, 2016 Unit 2 standby liquid control system on June 20, 2016 The inspectors selected these systems based on their risk significance relative to the reactor safety cornerstones at the time they were inspected. The inspectors reviewed applicable operating procedures, system diagrams, the UFSAR, TSs, WOs, IRs, and the impact of ongoing work activities on redundant trains of equipment in order to identify conditions that could have impacted the system's performance of their intended safety functions. The inspectors also performed field walkdowns of accessible portions of the systems to verify system components and support equipment were aligned correctly and were operable. The inspectors examined the material condition of the components and observed operating parameters of equipment to verify that there were no deficiencies. The inspectors also reviewed whether Exelon staff had properly identified equipment issues and entered them into the CAP for resolution with the appropriate significance characterization.

b. Findings

No findings were identified.

.2 Full System Walkdown

Main article: IP 71111.04S

No findings were identified.

1R07 Heat Sink Performance

Main article: IP 71111.07A

a. Inspection Scope

The inspectors reviewed the Unit 2 'A' residual heat removal (RHR) heat exchanger (HX) 2RHS*E1A to determine its readiness and availability to perform its safety functions. The inspectors reviewed the design basis for the component and verified Exelon's commitments to NRC Generic Letter 89-13, "Service Water System Requirements Affecting Safety-Related Equipment." The inspectors performed a visual inspection of the HXs and reviewed the results of previous inspections of the 'A' RHR HX 2RHS*E1A. The inspectors discussed the results of the most recent inspection with engineering staff and reviewed pictures of the as-found and as-left conditions. The inspectors verified that Exelon initiated appropriate corrective actions for identified deficiencies. The inspectors also verified that the number of tubes plugged within the HX did not exceed the maximum amount allowed in the design basis.

b. Findings

No findings were identified.

1R08 Inservice Inspection

Main article: IP 71111.08

a. Inspection Scope

From April 18 to 22, 2016, the inspectors conducted an inspection and review of inservice inspection (ISI) activities in order to assess the effectiveness of Exelon's program for monitoring degradation of the reactor coolant system (RCS) boundary, risk-significant piping boundaries, and the containment system boundaries during the Unit 2 15th refueling outage.

Nondestructive Examination (NDE) and Welding Activities (IP Section 02.01) The inspectors observed a sample of in-process NDE, reviewed completed documentation, and interviewed Exelon personnel to determine if the NDE activities performed as part of the third interval, third period, of the Unit 2 ISI program were conducted in accordance with the requirements of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code,Section XI, 2004 Edition with no addenda. For augmented examinations, the inspectors verified that activities were performed in accordance with Exelon's augmented inspection program and procedures and with any applicable industry guidance documents. The inspectors verified that indications and defects, if present, were dispositioned in accordance with the ASME Code or an NRC-approved alternative and verified that relevant indications were compared to previous examinations to determine if any changes had occurred. Activities included a review of ultrasonic testing (UT), radiographic testing (RT), and visual testing (VT). The inspectors reviewed certifications of the NDE technicians performing the examinations and verified that the inspections were performed in accordance with qualified NDE procedures and industry guidance. For UT activities, the inspectors also verified the calibration of equipment used to perform the examinations. The inspectors verified that the test results were reviewed and evaluated by certified Level III NDE personnel and that the parameters used in the test were in accordance with the limitations, precautions, and prerequisites specified in the test procedure. ASME Code Required Examinations Direct observation of the manual UT of the RHR 'A' HX head-to-shell weld (2RHS*E1A HW100A).

Direct observation of the manual UT of MOV40A valve-to-pipe weld (2RHS*66-19-FW004) in the RHR system.

Documentation review of the automated phased array UT of the N6C nozzle safe end-to-safe end extension weld (2RPV-KC26) in the RHR system.

Documentation review of the RT of two slip-on flange welds (FW-01 and FW-02 in 2-CPS-012-24-2), performed as part of a modification activity in the containment purge system.

Documentation review of the VT of the containment (i.e., drywell and suppression pool) interior penetrations and surfaces. The inspectors independently examined the condition of the drywell and suppression pool liner surfaces at all accessible floor elevations and compared those documented exams to the inspector walkdowns. Other Augmented, License Renewal or Industry Initiative Examinations The inspectors conducted direct observation of the remote enhanced VT records of the reactor vessel internals during in-vessel visual inspection activities to determine if they were in accordance with BWRVIP-03. Specifically, the inspectors reviewed the core shroud welds and jet pump components including the main wedges and the wedge hold down system modification.

Review of Previous Indications The inspectors did not review any previous indications because there were no relevant indications from the previous refueling outage that required evaluation for continued service at this time.

Welding on Pressure Boundary Systems The inspectors reviewed the pressure boundary risk-significant welding activity, including the associated NDE, of two slip-on flange-to-pipe welds (FW-01 and FW-02 in 2-CPS-012-24-2) as part of a modification to the containment purge system.

Specifically, the scope of the activity was to relocate two suppression chamber exhaust isolation valves, AOV-109 and AOV-111, which was part of the containment hardened vent modification. The inspectors performed a documentation review of the welding activities conducted before the outage to determine if the welding, RT/VT examinations, and final acceptance were performed in accordance with the ASME code requirements.

The inspectors reviewed the weld procedure specification to ensure it contained the required essential and supplemental essential weld variables and that those variables were within the ranges demonstrated by the supporting qualification record. The modification was performed under WO C93017630.

Identification and Resolution of Problems (IP Section 02.05)

The inspectors reviewed a sample of Unit 2 corrective action reports, which identified NDE indications, deficiencies, and other non-conforming conditions since the previous refueling outage and during the current outage. The inspectors verified that non-conforming conditions were properly identified, characterized, evaluated, and that corrective actions were identified and entered into the corrective action program for resolution.

b. Findings

No findings were identified.

1R11 Licensed Operator Requalification Program and Licensed Operator Performance

Main article: IP 71111.11Q

.1 Quarterly Review of Licensed Operator Requalification Testing and Training (2 samples)

a. Inspection Scope

The inspectors observed: Unit 1 licensed operator simulator training scenario, which involved a drifting control rod, a loss of a recirculation pump, fuel damage, and an emergency condenser tube leak on May 31, 2016 Unit 2 licensed operator simulator training scenario, which involved a loss of a condensate booster pump, a tornado watch with high wind warning, a partial loss of offsite power, and a loss-of-coolant accident inside containment on June 7, 2016 The inspectors evaluated operator performance during the simulated event and verified completion of risk-significant operator actions, including the use of abnormal and emergency operating procedures. The inspectors assessed the clarity and effectiveness of communications, implementation of actions in response to alarms and degrading plant conditions, and the oversight and direction provided by the control room supervisor. The inspectors verified the accuracy and timeliness of the emergency classifications made by the shift manager and the TS action statements by the unit supervisor. Additionally, the inspectors assessed the ability of the crew and training staff to identify and document crew performance problems.

b. Findings

No findings were identified.

.2 Quarterly Review of Licensed Operator Performance in the Main Control Room

(2 samples)

The inspectors reviewed operability determinations for the following degraded or non-conforming conditions based on risk significance of the associated components and systems: Unit 1 hydraulic control unit following discovery of foot valve leakage on April 8, 2016 Unit 2 service water pipe through-wall leak on reactor building elevation 240 feet on April 21, 2016 Unit 2 high pressure core spray operability during reactor startup on May 9, 2016 Unit 2 directional control valve cap screw susceptibility to failure based General Electric-Hitachi Service Information Letter 678 on May 13, 2016 Unit 1 EDG 102 and EDG 103 jacket water cooling tank couplings on May 13, 2016 Unit 1 stack radiation monitor RAM-112-07A failed downscale on May 23, 2016 Unit 2 RCIC tuning at 150 psig reactor pressure on June 7, 2016 Unit 2 EDG 3 exhaust damper on June 7, 2016 Unit 2 reactor pressure vessel (RPV) cold reference legs missed preventative maintenance on June 7, 2016 Unit 2 main steam isolation valve 2MSS*AOV7D limit switch 33-5 failure on June 13, 2016 Unit 1 door D109, control room envelope boundary and high energy line break barrier, found ajar on June 22, 2016 Unit 1 diesel fuel storage tank particulate levels on June 23, 2016 The inspectors evaluated the technical adequacy of the operability determinations to assess whether 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 Exelon's evaluations to determine whether the components or systems were operable. The inspectors confirmed, where appropriate, compliance with bounding limitations associated with the evaluations. Where compensatory measures were required to maintain operability, such as in the case of operator workarounds (OWAs), the inspectors determined whether the measures in place would function as intended and were properly controlled by Exelon. Based on the review of the selected OWAs listed above, the inspectors verified that Exelon identified OWAs at an appropriate threshold and addressed them in a manner that effectively managed OWA-related effects on operators and SSCs.

b. Findings

Introduction.

The inspectors identified a Green NCV of Unit 2 TS 3.5.1, "ECCS Operating," and TS 3.5.3, "RCIC System," for failure to ensure all necessary attendant instrumentation required for the systems to perform their specified safety functions were capable of performing their related support function in all required modes of applicability.

Specifically, inspectors identified the Unit 2 wide range level indication was inaccurate during Mode 2 with reactor vessel pressure at 200 psig, a mode of applicability requiring both HPCS and RCIC to be operable. Due to the instrument inaccuracies at lower pressures and temperatures, a false high-high water level trip signal was generated preventing both HPCS and RCIC from being actuated automatically or manually; therefore, rendering them inoperable.

Description.

The HPCS system is designed to provide and maintain adequate inventory inside the RPV to limit fuel cladding temperatures in the event of breaks in the reactor coolant pressure boundary. The system is automatically initiated on a high drywell pressure signal or low-low water level signal. It is designed to operate independently of all other systems over the entire range of pressure differences from greater than normal operating pressure to zero in modes 1, 2, and 3. The RCIC system designed to supply makeup water to the reactor vessel when the reactor is in a hot shutdown condition and is isolated from the main condenser with the reactor feedwater system not in operation. On May 5, 2016, during a reactor startup from a refueling outage with reactor pressure at approximately 200 psig, the inspectors questioned why the high water level annunciator was still lit, along with the high water level seal-in white light bulb for the HPCS system. The inspectors noted that the reactor vessel water level was at its normal level, and no longer in a high water level condition as it was earlier for hydro testing. Operators provided procedure N2-OP-101A, "Reactor Startup," Revision 02600, Step 2.44, which states, "prior to RPV pressure exceeding 600 psig, perform the following:

2.44.1 - Reset HPCS reactor high water level at Panel P601 2.44.2 - Observe HPCS reactor high level seal-in white light extinguishes" Inspectors questioned why the alarm would be on until 600 psig given HPCS was required to be operable before entering Mode 3, "Hot Shutdown."

The RPV water level signal for HPCS and RCIC comes from the wide range level indication. For HPCS, once water level reaches the high water level condition, the HPCS injection valve automatically shuts to prevent overfilling the reactor. The HPCS pump will continue to run until the operator takes action. If RPV water level reaches the low-low level condition, the injection valve will automatically reopen. The inspectors' review of the wide range level calibration calculation CS-ISC*05, "Reactor Water Level -

Off Normal Calibration Conditions," Revision 1, showed that until reactor pressure reaches approximately 600 psig, the wide range level indication reads higher than actual water level. This difference is a linear function affected by the water temperature and pressure of the reactor vessel. By not having accurate level instrumentation to support HPCS and RCIC initiation during their respective TS Mode of Applicability, both systems should have been declared inoperable and a reactor startup should not have commenced.

Additionally, by the wide range level indication not being accurate at the low reactor pressure and temperature conditions, this would prevent the respective injection valves from automatically opening on a high drywell pressure signal, and also prevent operators from initiating each system using the manual push button. This would require operators to manually open the HPCS and RCIC injection valves during these conditions should a loss of offsite power or loss-of-coolant accident occur. Therefore, the safety function to inject high pressure water into the reactor vessel would have been delayed but not lost.

The impact of the wide range level indication not being accurate should have resulted in Unit 2 entering the following limiting condition for operation (LCO) action statements for inoperable equipment based on the reactor startup conditions specified. The LCOs include the following: LCO 3.3.5.1 "ECCS Instrumentation" Conditions A, B, and H LCO 3.3.5.2 "RCIC Instrumentation" Conditions A, B, and E LCO 3.3.3.1 "PAM Instrumentation" Condition A LCO 3.5.1 "ECCS - Operating" Conditions B and D LCO 3.5.3 "RCIC System" Conditions A and B Since HPCS and RCIC support each other, it was required by TSs that operators exit the Mode of Applicability, i.e. shutdown. The Mode of Applicability for HPCS is Mode 1, 2 and 3. The Mode of Applicability for RCIC is Mode 1, 2, and 3 with reactor steam dome pressure greater than 150 psig. The inspectors did discover that prior to conversion to Improved Standard Technical Specifications (ISTS), Nine Mile Point Unit 2 Technical Specification Amendment 73 contained a note in TS 3.3.3," ECCS Instrumentation," that addressed the known hot calibration/cold operation instrument inaccuracy and did not require automatic or manual injection function to be operable when indicated level was great than level 8 coincident with reactor pressure less than 600 psig. However, this note was not carried forward during the conversion to the ISTS, and is not a part of the current operating license.

Exelon generated IR 02667837 to address the inspectors' concern regarding the wide range level indication. An action has been created to evaluate the impact of the wide range level discrepancy with regard to its impact on safety related functions to supply water in their TS Mode of Applicability. Exelon also plans to assess the need for a TS amendment.

Analysis.

Exelon's failure to ensure operability of all necessary instrumentation used to support HPCS and RCIC in accordance with TS 3.5.1 and TS 3.5.3 is a performance deficiency that was reasonably within Exelon's ability to foresee and prevent. This performance deficiency was more than minor because it was associated with the Equipment Performance attribute of the Mitigating Systems cornerstone and affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, Exelon failed to recognize that the wide range level indication did not provide accurate indication at low reactor pressures and temperatures, preventing automatic safety-related functions associated with high drywell pressure automatic initiation signals and manual start functions rendering those systems inoperable in a mode were they are required.

In accordance with IMC 0609.04, "Initial Characterization of Findings," and Exhibit 2 of IMC 0609, Appendix A, "The Significance Determination Process for Findings At-Power,"

issued June 19, 2012, the inspectors determined that the finding was of very low safety significance (Green), because the finding was not a design or qualification deficiency, did not represent a loss of system safety function, and did not screen as potentially risk significant due to seismic, flooding, or severe weather initiating event.

The finding had a cross-cutting aspect in the area of Problem Identification and Resolution, Identification. Exelon had many opportunities, including during the reactor startup in May of 2016, to question operability of the instrumentation and supported safety systems that are provided input for automatic initiation and isolation signals. However, Exelon personnel failed to identify that the wide range level indication did not support operability of the HPCS and RCIC systems during reactor startup on May 5, 2016, based on Nine Mile Point's current operating license. [P.1]

Enforcement.

TS 3.5.1, "ECCS - Operating" requires HPCS to be operable in Mode 1, 2, and 3. TS 3.5.3, "RCIC System" requires RCIC to be operable in Mode 1, 2 and 3 with reactor steam dome pressure greater than 150 psig. Contrary to the above, the HPCS and RCIC systems were not operable during their required Mode of Applicability.

Specifically, on May 5, 2016, inspectors identified that the wide range level indication was not accurate until the reactor reached approximately 600 psig during startup, resulting in the HPCS system not being operable during Mode 2 and 3 prior to the reactor reaching approximately 600 psig, and RCIC not being operable during Mode 2 and 3 between reactor pressures 150 psig and 600 psig. Because the violation was of very low safety significance (Green) and was entered into Exelon's CAP as IR 02667837, this violation is being treated as an NCV, consistent with Section 2.3.2 of the NRC Enforcement Policy. (NCV 05000410/2016002-02, Failure to Identify Wide Range Level Indication Impacts Operability of HPCS and RCIC)

1R18 Plant Modifications

Main article: IP 71111.18

.1 Temporary Modifications (1 sample)

a. Inspection Scope

The inspectors reviewed the Unit 2 Engineering Change Package (ECP) 15-000699, Evaluation of Smoke Detection Capability in the NMPNS 2 Relay Room, on March 30, 2016 to determine whether the modifications affected the safety functions of systems that are important to safety. The inspectors reviewed 10 CFR 50.59 documentation and post-modification testing results, and conducted field walkdowns of the modifications to verify that the temporary modifications did not degrade the design bases, licensing bases, and performance capability of the affected systems.

b. Findings

No findings were identified.

.2 Permanent Modifications (4 samples)

a. Inspection Scope

b. Findings

No findings were identified. Cornerstone: Emergency Preparedness

1EP6 Drill Evaluation

Main article: IP 71114.06

Training Observations

a. Inspection Scope

The inspectors observed a simulator training evolution for Unit 2 licensed operators on June 3, 2016, which required emergency plan implementation by an operations crew.

Exelon planned for this evolution to be evaluated and included in performance indicator (PI) 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 if there were any weaknesses and deficiencies in the crew's performance and ensure that Exelon evaluators noted the same issues and entered them into the CAP.

b. Findings

No findings were identified.

RADIATION SAFETY

Cornerstone:

Public Radiation Safety and Occupational Radiation Safety

2RS1 Radiological Hazard Assessment and Exposure Controls

Main article: IP 71124.01

a. Inspection Scope

During April 18 to 22, 2016, the inspectors reviewed Exelon's performance in assessing and controlling radiological hazards in the workplace. The inspectors used the requirements contained in 10 CFR Part 20, TSs, applicable regulatory guides (RGs), and the procedures required by TSs as criteria for determining compliance.

Inspection Planning

The inspectors reviewed the performance indicators for the occupational radiation safety cornerstone, radiation protection program audits, and reports of operational occurrences in occupational radiation safety since the last inspection.

Radiological Hazard Assessment (1 sample) The inspectors conducted independent radiation measurements during walkdowns of the facility and reviewed the radiological survey program, air sampling and analysis, continuous air monitor use, recent plant radiation surveys for radiological work activities, and any changes to plant operations since the last inspection to determine survey adequacy of any new radiological hazards for onsite workers or members of the public.

Instructions to Workers (1 sample)

The inspectors reviewed high radiation area work permit controls and use, observed containers of radioactive materials, and assessed whether the containers were labeled and controlled in accordance with requirements.

The inspectors reviewed several occurrences where a worker's electronic personal dosimeter alarmed. The inspectors reviewed Exelon's evaluation of the incidents, documentation in the CAP, and whether compensatory dose evaluations were conducted when appropriate. The inspectors verified follow-up investigations of actual radiological conditions for unexpected radiological hazards were performed. Contamination and Radioactive Material Control (1 sample)

The inspectors observed the monitoring of potentially contaminated material leaving the radiological controlled area and inspected the methods and radiation monitoring instrumentation used for control, survey, and release of that material. The inspectors selected several sealed sources from inventory records and assessed whether the sources were accounted for and were tested for loose surface contamination. The inspectors evaluated whether any recent transactions involving nationally tracked sources were reported in accordance with requirements.

Problem Identification and Resolution (1 sample) The inspectors evaluated whether problems associated with radiation monitoring and exposure control (including operating experience) were identified at an appropriate threshold and properly addressed in the CAP.

b. Findings

Introduction.

A Green self-revealing NCV of TS 5.4.1, "Procedures," was identified when a worker performed a significant radiological work activity without notifying radiation protection and, as a result, did not comply with procedure RP-AA-1008 "Unescorted Access to and Conduct in Radiologically Controlled Areas," Revision 5, in being briefed on the radiological work conditions and requisite radiological controls associated with performance of the Unit 2 reactor seal cleaning work activity.

Description.

On April 11, 2016, a worker entered the Unit 2 reactor cavity to perform inspection of the Unit 2 reactor seal that was highly contaminated. Although not previously discussed with radiation protection staff, the worker cleaned the highly contaminated reactor seal with rags, and carried the highly contaminated rags (5 rem/hr)out of the reactor cavity in his hand and subsequently received an unplanned radiation exposure to his hand. Step 6.5.1.1 of procedure N2-MPM-GEN-903, "Reactor Vessel Disassembly," requires personnel to inspect and clean the reactor seal as necessary.

However, the specifics regarding this seal cleaning work step was not communicated to radiation protection personnel and as a result the radiation protection technicians who were overseeing the reactor cavity area did not anticipate this work activity. As a result, the worker who performed the task was not briefed with respect to the radiological work to be performed or the appropriate radiological controls for performing Step 6.5.1.1 of procedure N2-MPM-GEN-903. Upon entering the reactor cavity, the worker proceeded to inspect and clean the seal with rags, and carried the highly contaminated rags in his hand out of the reactor cavity when the work was completed.

Upon reaching the refueling floor, a continuous air monitor alarmed. Radiation protection staff responded to the alarm and took the rags from the worker. A follow-up survey of the rags indicated 5 rem/hr on contact which resulted in unintended, unnecessary exposure to the worker's hand. Actions were taken to control the rags and surveys of the refuel floor were conducted to mitigate the situation.

Analysis.

The failure of the worker to discuss the radiological work activity with radiation protection, and to be briefed by radiation protection on the radiological conditions and requisite radiological controls for the work activity is a performance deficiency that was reasonably within Exelon's ability to foresee and correct. The finding was determined to be more than minor because it is associated with the human performance attribute of the Occupational Radiation Safety cornerstone and affects the associated cornerstone objective to ensure the adequate protection of the worker health and safety from exposure to radiation from radioactive material during routine civilian nuclear reactor operation. Specifically, the worker did not inform radiation protection personnel of the scope of work to be completed, and as a result, appropriate radiological work condition briefings and radiological controls to protect the worker from unnecessary radiation exposure were not implemented. Using IMC 0609, Appendix C, "Occupational Radiation Safety Significance Determination Process," dated August 19, 2008, the finding was determined to be of very low safety significance (Green) because it did not involve: (1) ALARA occupational collective exposure planning and controls, (2) an overexposure, (3) a substantial potential for overexposure, or (4) an impaired ability to assess dose.

The cause of the finding is related to the cross-cutting aspect of Human Performance Team Work, in that individuals and work groups did not communicate and coordinate their activities within and across organizational boundaries to ensure nuclear safety is maintained. Specifically, the worker did not communicate to radiation protection staff the scope of reactor seal cleaning activity that was to be performed which resulted in radiation protection not prescribing sufficient radiological controls for this high contamination work activity and resulted in the unintended exposure to the worker's hand. [H.4].

Enforcement.

TS 5.4.1 written procedures be established, implemented, and maintained for those activities recommended in Regulatory Guide 1.33, Revision 2, Appendix A, February 1978. Regulatory Guide 1.33, Revision 2, Appendix A, February 1978 recommends, in part, procedures for access control to radiation areas including a radiation work permit system. Procedure RP-AA-1008 "Unescorted Access to and Conduct in Radiologically Controlled Areas," implements these requirements. Steps 4.1.5 and 4.1.6 require, in part, that workers know their job scope and location and understand the radiological conditions for the work activity prior to entry.

Contrary to the above, on April 11, 2016, a worker conducting inspection of the reactor seal performed cleaning of the reactor seal that was not included in the radiological briefing and was not provided the necessary radiological work controls for this work activity. Exelon's immediate corrective actions included reinforcing the need to properly communicate radiological work activities with radiation protection, and require workers to carry work orders with them to improve communications with radiation protection.

Because this finding was determined to be of low safety significance (Green) and was entered into Exelon's CAP (AR 02654591), this violation is being treated as an NCV consistent with Section 2.3.2 of the NRC Enforcement Policy. (NCV 00050410/2016002-03, Failure to Understand Radiological Conditions Result in Unintended Exposure)

2RS2 Occupational ALARA Planning and Controls

Main article: IP 71124.02

a. Inspection Scope

During April 18 to 22, 2016, the inspectors assessed Exelon's performance with respect to maintaining occupational individual and collective radiation exposures ALARA. The inspectors used the requirements contained in 10 CFR 20, applicable RGs, TSs, and procedures required by TSs as criteria for determining compliance.

Radiological Work Planning The inspectors selected the following radiological work activities based on exposure significance for review: Radiation Work Permit (RWP)16-512, 513, 514 Control Rod Drive Activities RWP 16-542 Drywell Permanent Shielding Modification RWP 16-524 Nozzle and Vessel Inservice Inspection and Associated Support RWP 16-520 Drywell Snubber Testing and Inspections For each of these activities, the inspectors reviewed ALARA work activity evaluations, exposure estimates, and exposure reduction requirements.

Verification of Dose Estimates and Exposure Tracking Systems The inspectors reviewed the current annual collective dose estimate, basis methodology, and measures to track, trend, and reduce occupational doses for ongoing work activities.

The inspectors evaluated the adjustment of exposure estimates, or re-planning of work.

Radiation Worker Performance (1 sample)

The inspectors observed radiation worker and radiation protection technician performance during radiological work to evaluate worker ALARA performance according to specified work controls and procedures. Workers were interviewed to assess their knowledge and awareness of planned and/or implemented radiological and ALARA work controls.

Problem Identification and Resolution (1 sample) The inspectors evaluated whether problems associated with ALARA planning and controls were identified at an appropriate threshold and properly addressed in the CAP.

b. Findings

No findings were identified.

2RS3 In-Plant Airborne Radioactivity Control and Mitigation

Main article: IP 71124.03

a. Inspection Scope

During April 18 to 22, 2016, the inspectors reviewed the control of in-plant airborne radioactivity and the use of respiratory protection devices in these areas. The inspectors used the requirements in 10 CFR Part 20, RG 8.15, RG 8.25, NUREG\CR-0041, TS, and procedures required by TS as criteria for determining compliance.

Inspection Planning

The inspectors reviewed the UFSAR to identify ventilation and radiation monitoring systems associated with airborne radioactivity controls and respiratory protection equipment staged for emergency use. The inspectors also reviewed respiratory protection program procedures and current performance indicators for unintended internal exposure incidents.

Engineering Controls (1 sample)

The inspectors reviewed operability and use of both permanent and temporary ventilation systems, and the adequacy of airborne radioactivity radiation monitoring in the plant based on location, sensitivity, and alarm set-points.

b. Findings

No findings were identified.

OTHER ACTIVITIES

4OA1 Performance Indicator Verification

Main article: IP 71151

.1 Reactor Coolant System Specific Activity and Reactor Coolant System Leak Rate

The inspectors identified one trend concerning quality of documentation in Exelon's CAP. This trend had previously been identified by the inspectors and was entered into Exelon's CAP as IR 2620310. Since documentation of this trend the inspectors have identified additional issues that have not been documented or have not been documented in a timely manner per Exelon's procedure PI-AA-120, Revision 6, "Issue Identification and Screening Process." In both cases discussed below Exelon's failure to document the issues was considered a minor procedural violation of Exelon's procedure PI-AA-120: IR 02620306 was written on February 1, 2016 in response to the inspectors identifying eight issue reports that were either not entered into the CAP in a timely manner or were not coded appropriately as NRC-identified. Each issue was reviewed by the inspectors to determine if any non-compliance reached a more than minor threshold. No issues of more than minor threshold were identified. Exelon initiated a crew clock reset for Engineering, Operations, and Regulatory Affairs Departments. The IRs were properly updated or generated in response to the inspectors concerns.

IR 2678327 was written after the inspectors questioned Exelon about the status of a post maintenance test (PMT) after replacement of the half load solenoid for the 11 instrument air compressor. The original PMT acceptance criteria as stated in the work order (C92984760) was no leakage. During conduct of the PMT leakage was noted. No CAP entry was made documenting the failed PMT. PI-AA-120, "Issue Identification and Screening Process," Revision 6, Section 4.3 directs issue origination using guidance in Attachment 4. Attachment 4, Section 1.22 identified "Failed Post Maintenance test." Contrary to this, the inspectors identified that the PMT was not met, and an issue report was not written in accordance with PI-AA-120.

The inspectors determined this issue was minor because the system engineer subsequently performed an evaluation to determine if the amount of leakage for the solenoid could still meet acceptance criteria in accordance with MA-AA-716-010-1103, "Fluid Sealing Technology Program," Revision 3 for the PMT and found the results to meet the criteria.

IR 2682226 was written after the inspectors questioned Exelon about the completion of a step in the work order to perform a Residual Heat Removal (RHR) heat exchanger preventive maintenance (C91927861). The step required verification with the RHR system engineer that as-found performance monitoring data had been collected prior to cleaning beginning on the heat exchanger. When questioned by the inspector, Exelon staff including the RHR system engineer could not recollect the verification occurring, and were uncertain if the as-found data had been collected.

PI-AA-120, "Issue Identification and Screening Process," Revision 6, Section 4.3 directs issue origination using guidance in Attachment 4. Attachment 4, Section 2.10 identifies "Procedures found to provide incorrect direction that would cause plant or personnel risk if followed." Contrary to this, the inspectors identified that the step was not accurate, and an issue report was not written in accordance with PI-AA-120.

The inspectors determined the issue to be minor because a review of the heat exchanger history revealed the as-found performance monitoring data had been collected under a separate work order (C90716549), precluding impact on the plant and ensuring that heat transfer characteristics were captured prior to heat exchanger tube cleaning.

.3 Annual Sample: Failure of Unit 1 EDG 103 to Start on March 15, 2016

a. Inspection Scope

The inspectors performed an in-depth review of Exelon's response and corrective actions associated with the Unit 1 EDG 103 failure to start during monthly surveillance testing on March 15, 2016. The inspectors reviewed the ACE performed by Exelon staff to determine whether Exelon staff appropriately identified, characterized, and corrected problems associated with the issue and whether the planned or completed corrective actions were appropriate. The inspectors also reviewed IRs, procedures, drawings, engineering change documents, completed WOs associated with troubleshooting, past work on the Woodward governor, and performed interviews of Exelon staff personnel.

b. Findings and Observations

No findings were identified.

Emergency AC power is supplied by two electromotive division (EMD) locomotive diesel generators, EDG 102 and EDG 103. Fuel oil input to each cylinder is controlled by a Woodward governor, model UG-8. In addition to the Unit 1 EDGs, the Unit 2 Division 3 EDG also employs the use of the Woodward governor, model UG-8. The UG-8 governor contains a shutdown solenoid which is energized to shutdown the EDG by repositioning the fuel rack to the no fuel position. On March 15, 2016, Exelon staff attempted to perform a monthly surveillance test of EDG 103. However, when the manual start signal was applied, the EDG failed start. Exelon troubleshooting determined the potential failure mechanism to be the shutdown solenoid. Exelon's immediate corrective action for EDG 103 failing to start was to replace the Woodward governor top hat, and its shutdown solenoid.

Exelon performed an ACE of the Unit 1 EDG 103 failure to start on March 15, 2016, under IR 02640675. Exelon determined the cause to be mechanical binding in the shutdown solenoid. Exelon performed a failure of the analysis which determined some resistance to movement from a burr and some debris. Because the shutdown solenoid was stuck in the down position, the governor maintained the fuel rack to the EDG in a position to restrict fuel injection. Exelon also performed an extent of condition run of EDG 102, in accordance with TS, on March 15, 2016, to ensure a similar condition would not prevent it from running. EDG 102 performed its run successfully, ensuring it remained operable. Although an extent of condition run was not performed for the Unit 2 Division 3 EDG, which has the same model Woodward governor valve, the EDG was run successfully on February 24, 2016. Exelon plans future replacements for Unit 1 EDG 102 and Unit 2 Division 3 EDG governor shutdown solenoids, which are being tracked as corrective actions (CAs) in IR 02640675. Exelon also modified its preventive maintenance (PM) to be performed at an eight-year frequency as opposed to the previous frequency of 10 years.

Preventive Maintenance Strategy The inspectors reviewed past PM practices associated with the UG-8 model Woodward governor at NMPNS. Prior to the license transfer of NMPNS from Constellation Energy to Exelon 2014, NMPNS did not implement a routine PM or replacement strategy for the governor. Although no PM was in place, based on operating experiences, NMPNS replaced its governors in 2005 with an energize to shutdown solenoid. Prior to 2005, the governors were energized to run, and were having repeat failures throughout the industry. Between 2005 and 2016 no additional PM or replacement occurred for the governor. The inspectors reviewed PM practices recommended by the vendor and industry and determined Exelon's current and past practices appeared to be in alignment with the vendor's recommendations.

Operating Experience The inspectors also performed a review of applicable operating experience regarding Woodward governor UG-8 models and shutdown solenoids. The inspectors noted that Exelon identified a similar failure at the LaSalle County Station, as documented in IR 00828185 on October 8, 2008. The inspectors' review of the issue determined that no operating experience communication was received from LaSalle regarding this event.

There was no requirement for such an event to be communicated to the sites. Without receiving information regarding the potential failure mechanism of the shutdown solenoid, Exelon staff were not able to enter the issue into their CAP at NMPNS to ensure potential corrective actions were put in place. The inspectors also performed a review of NRC internal operating experience and did not identify any events that would have been applicable to NMPNS regarding the potential for the shutdown solenoid to become stuck and prevent EDG operation. The inspectors determined that Exelon staff appropriately considered operating experience for applicability in accordance with PI-AA-115, "Operating Experience Program," Revision 1.

Past Operability and Reportability The inspectors identified during review of the ACE that Exelon did not appropriately consider past operability. Exelon generated IR 02671608 to document the need for a formal past operability review. PI-AA-125, Corrective Action Program, Revision 4, Section 4.1.2 states, "if at any time (e.g., during an investigation, review of a CA closure, review of a previous CR) a SCAQ or CAQ or any question of either current or past operability/reportability arises, then initiate an Issue Report (IR) in accordance with PI-AA-120, 'Issue Identification and Screening Process,'" which was not performed during the apparent cause investigation. Exelon performed a human performance review board (HURB) and determined that although the apparent cause template does not require a past operability review, the section created for a past operability review was inadequate.

Exelon generated an action item to re-assess past operability and reportability. Exelon developed a technical evaluation, documented in ECP-16-000443. Exelon determined through a review of design basis requirements, actual relay performance, worst case drift performance, and worst case EDG performance that although EDG 103 failed to start within 10 seconds, it would have started automatically and supported power to core spray to ensure full flow to the core within 60 seconds and meet the fuel peak clad centerline criteria of 2,200°F. Inspectors reviewed the relay start circuit for EDG 103 as well as the actual testing relay performance data. The inspectors' review determined that it was reasonable that on a second start of EDG 103, enough force would likely have caused the solenoid to become free and allow the fuel rack to reposition and support a second startup. Based on this conclusion, the inspectors determined that a reportable condition did not exist for EDG 103, because although the EDG failed to start on March 15, 2016, if called upon during a loss of offsite power EDG 103 would likely have started and provided power to core spray to inject in the core within the 60-second time limit to ensure the fuel peak cladding temperature did not exceed its limit.

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

Main article: IP 71153

.1 Plant Events (1 sample)

a. Inspection Scope

For the plant events listed below, the inspectors reviewed and/or observed plant parameters, reviewed personnel performance, and evaluated performance of mitigating systems. The inspectors communicated the plant event(s) to appropriate regional personnel, and compared the event details with criteria contained in IMC 0309, "Reactive Inspection Decision Basis for Reactors," for consideration of potential reactive inspection activities. As applicable, the inspectors verified that Exelon made appropriate emergency classification assessments and properly reported the event in accordance with 10 CFR Parts 50.72 and 50.73. The inspectors reviewed Exelon's follow-up actions related to the events to assure that Exelon implemented appropriate corrective actions commensurate with their safety significance.

Unit 2 intermittent loss of Division I position indication during remote shutdown panel testing on May 4, 2016

b. Findings

No findings were identified.

.2 (Closed) Licensee Event Report (LER) 05000220/2014-002-00:

Unanalyzed Condition Due to Unfused Motor-Operated Valve Control Circuit (1 sample)

On May 8, 2014, during a review of industry operating experience, Exelon identified that an unfused control circuit associated with the Unit 1 reactor water cleanup isolation valve 12 could short circuit due to a fire in the circuit cable routing. Exelon postulated that fire damage to an unprotected DC circuit could cause the cable to self-heat and cause secondary fires along the associated cable route and, as a consequence, adversely affect equipment necessary to achieve and maintain safe shutdown. Exelon determined the cause of the condition was an original design configuration that did not conform to the current fire protection standards. There was no actual safety consequence for this condition. Exelon entered this issue into their CAP as condition report (CR)-2014-004630, implemented compensatory measures for the affected fire areas, and subsequently corrected this condition by the addition of fuses to the affected control circuit.

The inspectors reviewed the LER, the risk evaluation, corrective actions, drawings, associated WO, and the ECP. The inspectors documented the enforcement aspects of the identified technical issue below. The inspectors did not identify any new issues during the review of this LER. This LER is closed.

Introduction.

Exelon identified a violation of low to moderate safety significance (White) of Unit 1 Operating License Condition 2.D.(7) for failure to implement and maintain all aspects of the approved Fire Protection Plan (FPP). Specifically, Exelon did not ensure that one train of equipment to achieve and maintain post-fire safe shutdown would remain free of fire damage. This issue was evaluated by the inspectors and determined to warrant enforcement discretion.

Description.

In 2014, Exelon was conducting reviews to support of their transition to National Fire Protection Association Standard 805, "Performance-Based Standard for Fire Protection for Light Water Reactor Electric Generating Plants," for Nine Mile Point Unit 1. On May 8, 2014, during a review of industry operating experience, Exelon identified that an unfused control circuit associated with the Unit 1 reactor water cleanup isolation valve 12 could short circuit due to a fire in the circuit cable routing. Exelon postulated that fire damage to an unprotected DC circuit could cause the cable to self-heat and cause secondary fires along the associated cable route and, as a consequence, adversely affect equipment necessary to achieve and maintain safe shutdown. This deficiency effected several fire areas. Exelon determined the cause of the condition was an original design configuration that did not conform to the current fire protection standards. There was no actual safety consequence for this condition.

Exelon entered this issue into their CAP as condition report (CR)-2014-004630, implemented compensatory measures for the affected fire areas, reported the condition to the NRC in accordance with 10 CFR 50. 72 and 10 CFR 50.73, and subsequently corrected this condition by the addition of fuses to the affected control circuit.

Analysis.

Exelon's failure to identify a non-conforming condition which could adversely impact multiple fire areas and challenge the plant's ability to achieve and maintain safe shutdown was a performance deficiency that was reasonably within Exelon's ability to foresee and correct. This performance deficiency was licensee identified during transition to National Fire Protection Association Standard 805 and 10 CFR 50.48(c).

This performance deficiency was more than minor because it was associated with the Protection Against External Factors (e.g., fire) attribute of the Mitigating Systems cornerstone and adversely affected the cornerstone objective to ensure the availability and reliability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage).

Exelon evaluated this issue with NMPNS fire probabilistic risk assessment analysis and determined that the change in core damage frequency attributed to this issue was estimated to be 9.8E-6 per reactor year. A review of the dominant scenarios indicated that the risk from a large early release was bounded by the core damage assessment. A Region I senior reactor analyst reviewed Exelon's evaluation and concluded that the risk estimate was bounded by conservative assumptions and that this issue would be of no greater than low to moderate safety significance (White).

The inspectors did not assign a cross-cutting aspect to this finding because the performance deficiency occurred greater than three years ago; therefore, it is not considered to be indicative of current plant performance. This was an original plant design configuration issue that did not conform to current license fire protection standards.

Enforcement.

Unit 1 License Condition 2.D.(7), in part, requires Exelon to implement and maintain in effect all fire protection features described in licensee submittals and as approved by the NRC. The FPP required Exelon to maintain one train of equipment necessary to achieve and maintain safe shutdown free of fire damage. Contrary to the above, from 1969 (original construction) until March 28, 2015, Exelon did not implement and maintain in effect all fire protection features described in licensee submittals and as approved by the NRC. Specifically, Exelon postulated that fire damage to an unprotected DC circuit could cause the cable to self-heat and cause secondary fires along the associated cable route and, as a consequence, would adversely affect equipment necessary to achieve and maintain safe shutdown. Exelon entered this issue into their CAP as CR-2014-004630, and subsequently corrected the condition.

Exelon was in transition to National Fire Protection Association Standard 805 and 10 CFR 50.48(c) in accordance with a current License Amendment Request, the issue was identified by the licensee as part of a voluntary initiative to adopt 10 CFR 50.48(c), the violation was corrected, routine efforts such as normal surveillances or quality assurance activities would not have likely identified the violation, and the violation was not willful.

Therefore, the criteria established in NRC Enforcement Policy Section 9.1 and IMC 0305 Section 11.05 were met. Because all the criteria were satisfied, the NRC exercised enforcement discretion and did not issue a violation for this issue.

.3 (Closed) LER 05000410/2015-003-01:

Primary Containment Isolation Function for Some Valves Not Maintained During Surveillance Testing (1 sample) This LER was revised on March 31, 2016, to document the affect the subject issue had on the Safety System Functional Failure PI. On June 23, 2015, Exelon identified two separate instances where the Unit 2 primary containment isolation function was not maintained for Divisions I and II of the reactor vessel low water level, level 2, associated with reactor building closed loop cooling and reactor water cleanup primary containment isolation valves during the performance of surveillance procedure N2-ISP-ISC-Q005, "Quarterly Functional Test of Reactor Vessel Level 2 and Level 1 Instrument Channels," Revision 00600. N2-ISP-ISC-Q005 is used to test functionality of the Rosemount Analog Trip Systems for several isolation functions associated with reactor vessel water level 2 and reactor vessel water level 1 to satisfy TS 3.3.6.1.3-1.a. In preparation for these surveillance tests, Exelon inadvertently disabled the opposite division isolation function of the one being tested and when they performed the surveillances both primary containment isolation function divisions on reactor vessel low water level were lost.

Exelon determined the cause to be a failure to adequately communicate and reinforce the use of technical human performance tools during work preparation activities in accordance with HU-AA-102, "Technical Human Performance Practices," Revision 007. These events were reportable under 10 CFR 50.73 (a)(2)(v)(C) as events or conditions that could have prevented the fulfillment of the safety function of structures or systems that are needed to control the release of radioactive material.

In the original LER, Exelon reported that the failure to maintain the low water level isolation function for the reactor building closed loop cooling and reactor water cleanup primary containment isolation valves did not constitute a Safety System Functional Failure (SSFF) in accordance with NEI 99-02, "Regulatory Assessment Performance Indicator Guideline," Revision 7. NEI 99-02 defines a SSFF as any 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. Exelon reached this conclusion based on the availability of the high drywell pressure isolation function which in the case of the design basis accident would generate an isolation signal in lieu of the low water level isolation function. Exelon failed to recognize other analyzed transients which would result in a low water level without generating high drywell pressure, therefore the loss of the low water level isolation function in those cases could have prevented primary containment isolation.

In the revised LER, Exelon recognized their inaccurate determination and correctly classified this event as a SSFF. Furthermore, Exelon has committed to update their PI data submission in the second quarter of 2016 to reflect the input to the SSFF PI. The enforcement aspects of this issue regarding the inaccurate PI data submission are discussed in NRC Integrated Inspection Report 05000410/2015003. The inspectors did not identify any new issues during review of this revised LER. This LER is closed.

.4 (Closed) LER 05000410/2016-001-00:

Secondary Containment Inoperable Due to Simultaneous Opening of Airlock Doors (1 sample) LER 05000410/2016-001-00 reported that on April 7, 2016, the Unit 2 reactor building was breached when workers opened both inner and outer airlock doors, SA-262-2 and SA-262-3, simultaneously. The door was subsequently shut within 5 seconds. The event resulted in operators entering and exiting TS 3.6.4.1, Condition C. During the time the airlock doors were open simultaneously, vacuum remained above the TS required 0.25 inches of vacuum water gauge. Exelon concluded that the wait time of 5 seconds before traversing the airlock was not long enough to reduce the risk of an event. Corrective actions included revising advisory signs at the airlock doors to have individuals pause for 10 seconds instead of 5 seconds prior to entering the airlock. This action was put in place to ensure individuals had enough time to clear the airlock before opening of the doors simultaneously could occur.

The inspectors' review documented that since 2014, there have been 10 events at NMPNS associated with simultaneous airlock doors being opened. These events include the following: Unit 2 on April 2, 2014 as documented in LER 05000410/2014-007-00 Unit 2 on April 2, 2014 as documented in LER 05000410/2014-007-01 Unit 1 on August 13, 2014 as documented in LER 05000220/2014-004-00 Unit 1 on October 16, 2014 as documented in LER 05000220/2014-005-00 Unit 1 on October 20, 2014 as documented in LER 05000220/2014-006-00 Unit 1 on February 11, 2015 as documented in LER 05000220/2015-001-00 Unit 1 on March 3, 2015 as documented in LER 05000220/2015-002-00 Unit 1 on August 5, 2015 as documented in LER 05000220/2015-003-00 Unit 1 on September 4, 2015 as documented in LER 05000220/2015-004-00 Unit 2 on April 7, 2016 as documented in LER 05000410/2016-001-00 The inspectors noted that corrective actions in the past included installation of cameras, installation of signs at the access doors to advise waiting before opening of the doors to ensure personnel have cleared the airlock, and the repair of magnetic interlocks. This event did not lead to the loss of secondary containment vacuum. As part of corrective actions for the events associated with the secondary containment airlock doors, Exelon has also submitted License Amendment Request (LAR) for Unit 1 and Unit 2 (MF6974 and MF6975) to allow momentary simultaneous opening during entry and exit and not require entry into TS 3.4.3 for Unit 1 and TS 3.6.4.1 for Unit 2. The LAR is currently under review by the NRC staff.

The inspectors reviewed the LER for accuracy, the adequacy of proposed and completed corrective actions, and the appropriateness of the extent-of-condition review.

No findings or violations of NRC requirements were identified. This LER is closed.

4OA6 Meetings, Including Exit

On May 11, 2016, the inspectors presented a debrief of the inspection results related to LER 05000220/2014-002 documented in Section

4OA3 to Mr. Brandon Varga,

Regulatory Assurance.

On July 13, 2016, the inspectors presented the inspection results to Mr. Peter Orphanos, Site Vice President, and other members of the NMPNS staff. The inspectors verified that no propriety information was retained by the inspectors or documented in this report.

ATTACHMENT:

SUPPLEMENTARY INFORMATION

KEY POINTS OF CONTACT

Licensee Personnel

P. Orphanos, Site Vice President

R. Kreider, Plant Manager

M. Busch, Director Site Operations

J. Gerber, Manager Site Chemistry, Environment and Radwaste

S. Homoki, Senior Engineer

P. Kehoe, Engineering Analyst

M. Khan, Senior Manager Engineering

B. Knowlton, Site Engineer

R. Kreider, Plant Manager

K. Kristensen, Regulatory Principle Engineer

M. Kunzwiler, Manager Site Security

D. Moore, Manager Regulatory Assurance

B. Scaglione, Manager Engineering

A. Sterio, Director, Site Engineering

D. Tulowiecki, Manager Site Radiation Protection

B. Varga, Regulatory Assurance

LIST OF ITEMS OPENED, CLOSED, DISCUSSED, AND UPDATED

Opened/Closed

05000410/2016002-01 FIN Ineffective Corrective Action Results in Water

Intrusion to Battery Switchgear Room (Section 1R01)

05000410/2016002-02 NCV Failure to Identify Wide Range Level Indication

Impacts Operability of HPCS and RCIC

(Section 1R15)

05000410/2016002-03 NCV Failure to Understand Radiological Conditions Results in Unintended Exposure (Section 2RS1)

Closed

05000220/LER-2014-002-00 LER Unanalyzed Condition Due to Unfused Motor Operated Valve Control Circuit (Section 4OA3)

05000410/LER-2015-003-01 LER Primary Containment Isolation Function for Some Valves Not Maintained During Surveillance Testing (Section 4OA3)

05000410/LER-2016-001-00 LER Secondary Containment Inoperable Due to Simultaneous Opening of Airlock Doors on

June 21, 2016 (Section 4OA3)

LIST OF DOCUMENTS REVIEWED

Section 1R01: Adverse Weather Protection

Procedures

LS-AA-129, NERC Reliability Standard Compliance Program, Revision 5

N2-MPM-GEN-AO16

N2-MPM-GEN-AO16 PROCEDURE TO WO#

C93043729 N2-OP-70, Station Electric Feed and 115kv Switchyard, Revision 01800

OP-AA-108-117, Protected Equipment Program, Revision 4

OP-NM-108-1-7-1002, Offsite Power Operations and Interface Revision 002

PI-AA-120, Issue Identification and Screening Process, Revision 6

PI-AA-125, Corrective Action Program, Revision 3

SOP-33.A,3 115kv Grid Disturbances, Revision

00200

SOP-70, Major Grid Disturbance, Revision 0300

WC-AA-107, Seasonal Readiness, Revision 16

Drawings

12177-EC-16B-9 Sheet 2, C.W. Discharge Lines Encasement Plan & Details, Revision 9

2177-EC-58CA-3, Foundation Plan El237'-0", 249'-0" & 250'-0" Normal Switchgear Building, Revision 3

EP-001B, Power Piping Drawing Circulating Water Piping, Revision 14

Issue Reports

02006312

02091644

02664534

02668574

Work Orders

C93043729 02-01974-00

Miscellaneous

Site Certification Letter for Summer Readiness dated May 12, 2016

Unit 2 USAR, Revision 21 Design Change 2S11239

Section 1R04: Equipment Alignment

Procedures

N1-OP-12 Liquid Poison System, Revision 02900

N2-OP-31, Residual Heat Removal System, Revision 03100

N2-OP-61B, Standby Gas Treatment System, Revision 01400

N2-OP-72, Standby and Emergency AC Distribution System, Revision 01401

N2-OP-100A, Standby Diesel Generators, Revision 01700 N2-OP-100B, HPCS Diesel Generator, Revision 01700

Issue Reports

02640675

Drawings

12177-EJ-ECCS-A-0, Emergency Core Cooling Systems Flow Diagram, Revision 0

PID-36A, Piping & Instrumentation Diagram Standby Liquid Control, Revision 25

PID-53A, Piping & Instrumentation Diagram Control Building Ventilation and Air Conditioning, Revision 25

PID-53B, Piping & Instrumentation Diagram Control Building Ventilation and Air Conditioning, Revision 17

PID-53D, Piping & Instrumentation Diagram Control Building Ventilation and Air Conditioning, Revision 14

PID-53E, Piping & Instrumentation Diagram Control Building Ventilation and Air Conditioning, Revision 17

PID-104, Piping & Instrumentation Drawing Standby Diesel Generator System, Revision 23

PID-104A, Piping & Instrumentation Diagram Standby Diesel Gen System, Revision 23

Section 1R05: Fire Protection

Procedures

EP-CE-111, Emergency Classification and Protective Action Recommendations, Revision 001

EPIP-EPP-02-EAL, Emergency Action Level Matrix Unit 2, Revision 22 N2-FPI-PFP-0201, Unit 2 Pre-Plans, Revision 3

N2-FPI-PFP-0201, Unit 2 Pre-Fire Plans, Revision 4.01

OP-NM-201-005, Firefighting, Revision 00200

Miscellaneous

Unit 2 UFSAR, Section 9.5

Unit 2 UFSAR, Revision 3

Unit 2 UFSAR, Revision 21

Section 1R06: Flood Protection Measures

Drawings

PID-66F, Piping & Instrumentation Diagram Miscellaneous Drains, Revision 10

Issue Reports

02670324

Miscellaneous

Unit 2 UFSARNUREG-1900 Vol. 2, Safety Evaluation Report Related to the License Renewal of Nine Mile Point Nuclear Station, Units 1 and 2

Section 1R07: Heat Sink Performance

Procedures

GAP-HSC-02, System Aging Inspection and Cleanness Controls, Revision 02000

N2-MPM-RHS-410, Residual Heat Removal Heat Exchanger P.M., Revision 00402 S-TDP-REL-0102, Service Water Heat Exchanger and Component Inspection Guide, Revision 03

Issue Reports

02657679

2658530

Work Orders

C91927861

Section 1R08: Inservice Inspection

Procedures

ER-AA-335-018, Visual Examination of ASME IWE Class MC and Metallic Liners of IWL Class CC Components, Revision 11

GEH-PDI-UT-1, PDI Generic Procedure for the Ultrasonic Examination of Ferritic Pipe Welds, Revision 9

GEH-VT-205, Procedure for Invessel Visual Inspection (IVVI) of BWR 5 RPV Internals, Revision

13 TOPAZ 03, Procedure for Encoded Phased Array Ultrasonic Examination of Dissimilar Metal Welds, Revision B

Drawings

ISI-66-19, ISI Weld and Pipe Support Identification Drawing, Revision 5

IWE-GEN-001, Sheet 1, Containment ISI IWE Inspection Drawing Liner General Arrangement, Revision 1

IWE-GEN-002, Containment ISI IWE Inspection Drawing Liner Rollout, Revision 3

Issue Reports

01987033

02006130

02038077

02654772

02655157

02657052

02657131

Miscellaneous

BOP-MT-16-009, Magnetic Particle Testing Data Report for

FW-01 and

FW-01 in 2-CPS-012-24-2, dated February 22, 2016

CNG-NMP2-ISI-003, Nine Mile Point Unit 2 Third Ten-Year Interval ISI Plan, Revision 03

CNG-NMP2-OAR-003-014, Owner's Activity Report for Unit 2 14th Refueling Outage, dated July

16, 2014 INR NMP2R15

IVVI-16-02, Jet Pump 4, 5, 6, 15, and 16 Wedge and Mandrel, dated April 17, 2016 INR NMP2R15

IVVI-16-04, Jet Pump 6 Sensing Line, dated April 17, 2016

INR NMP2R15

IVVI-16-07, Steam Dryer Tie Bars, dated April 18, 2016

ISI-UT-16-042, UT Data Report for Weld 2RHS*E1A HW100A, dated April 21, 2016

ISI-UT-16-044, UT Data Report for Weld 2RHS-66-19-FW004, dated April 22, 2016

ISI-VT-16-205-215, VT Data Report for Suppression Chamber Elevations 200' to 240', dated April

30, 2016

ISI-VT-16-218-221, VT Data Report for Suppression Chamber Elevations 175' to 200', dated April

21, 2016

ISI-VT-16-279-290, VT Data Report for Drywell Elevations 288' to 329', dated April 29, 2016

LTR-BWR-ENG-16-015, Westinghouse Recommendations for Use of Main Wedge Hold Downs on JP 4/5/6/15/16 Use As-Is for One Cycle of Operation, Revision 0 RT Data Report for

FW-01 and

FW-02 in 2-CPS-012-24-2, dated March 28, 2016

SPC-14-01-N, Special Process Control and Inspection Audit Report, dated November 17, 2014

Weld Log Record for 2-CPS-012-24-2

FW-01 and

FW-02 under

C93017630, dated April 18, 2016

Welding Procedure Specification

WPS-1-1-BA-101, Carbon to Carbon SWAW and GTAW, Revision 16

Work Orders

C92744622

C92744636 C93017630

Section 1R11: Licensed Operator Requalification Program and Licensed Operator Performance

Procedures

N1-OP-101A, Plant Start-Up, Revision 03600 N2-IPM-ICS-R010, RCIC Tuneup, Revision 00400

N2-OP-101C, Plant Shutdown, Revision 02700

N2-OP-101D, Power Changes, Revision 02300

N2-OSP-ICS-R002, RCIC System Flow Test, Revision 00400

Issue Reports

02676658

Work Orders

C93456586

Section 1R12: Maintenance Effectiveness

Procedures

S-MRM-REL-0102, Structural Monitoring Program, Revision 00800

RM-AA-101-1008, Processing and Storage of Records, Revision 8

RM-AA-101, Records Management Program, Revision 10

PI-AA-125, Corrective Action Program, Revision 4

ER-AA-321, Administrative Requirements for Inservice Testing, Revision 12

Issue Reports

02044041

2049498

02626504

02634808

02638701

02657828 Work Orders

C92746645

Miscellaneous

NMPNS-IST-001, Pump and Valve Inservice Testing Program, Revision 8

Unit 2 USAR, Revision 21

Section 1R13: Maintenance Risk Assessments and Emergent Work Control

Procedures

LS-AA-119, Fatigue Management and Work Hour Limits, Revision 12

OU-NMP-103-101, Shutdown Safety Management Program, Revision 100

WC-AA-101-1008, On Line Risk Management and Assessment, Revision 2

Issue Reports

02666663

Miscellaneous

Unit 2 Shutdown Safety Management Plan, Revision 0

Section 1R15: Operability Determinations and Functionality Assessments

Procedures

EP-CE-121-1004, Nine Mile Point Equipment Matrix, Revision 001 N1-ARP-F3, Annunciator Response Procedure Panel F3, Revision 00600 N1-ARP-H1, Control Room Panel H1, Revision 00900 N1-CSP-Q504, Quarterly Diesel Fuel Oil Sampling and Analysis, Revision 00701 N1-MPM-044-658, Hydraulic Control Unit Preventative Maintenance, Revision 01200

N1-OP-5, Control Rod Drive System, Revision 05400

N1-OP-50B, Process Radiation Monitoring System, Revision 01300

N2-ISP-CSH-R107, Operating Cycle Channel Calibration of HPCS Initiation on Reactor Vessel Water Level Low Low, Level 2 and Isolation on High Level 8 Instrumentation Channels, Revision 00500 N2-ISP-RPS-R107, Operating Cycle Calibration of Reactor Scram on MSIV Closure Instrument Channels, MSIV Position Indication, and MSIV Slow Closure Stroke Time, Revision 01002 N2-OP-101A, Plant Start-Up, Revision 03600 N2-OSP-CSH-Q@002, HPCS Pump and Valve Operability and System Integrity Test, Revision 00600 N2-OSP-CSH-R301, Logic System Functional Test of HPCS Injection Valve Closure on Reactor Vessel High Level 8 Trip, Revision 00203 N2-OSP-CSH-Q@002, HPCS Pump and Valve Operability and System Integrity Test, Revision 00600 N2-OSP-MSS-Q002, Main Steam Isolation Valve Partial Exercise Test and Functional Test of RPS Main Steam Isolation Valve Closure in Operating Condition 1 or 2, Revision 00700 N2-PM-A@0002, RPV Instrument Reference Leg Leakage Test, Revision 00001

Drawings

C-18016-C, Control Rod Drive P&ID, Revision 42 F-42199-C, Emergency Diesel Fuel Oil Storage Tanks Sections and Detail, Revision 8

F-42188-C, Emergency Diesel Fuel Oil Storage Tanks Surfacing, Drainage and Underground Conduit and Piping Plan, Revision 001

ISPT-28B-3, Nuclear Boiler and Process Instrumentation Pressure Testing Diagram, Revision 3 807E172TY Sheet 3, Elementary Diagram High Pressure Core Spray System, Revision 4

807E172TY Sheet 7, Elementary Diagram High Pressure Core Spray System, Revision 27

PID-30B, Piping and Instrumentation Diagram Control Rod Drive Hydraulic System, Revision 14

PID-28A-18, P&ID Nuclear Boiler and Process Instrumentation, Revision 18

PID-28B-8, P&ID Nuclear Boiler and Process Instrumentation, Revision 8

PID-28C-11, P&ID Nuclear Boiler and Process Instrumentation, Revision 11

PID-30A, P&ID Control Rod Drive Hydraulic System, Revision 24

Issue Reports

02501333

02593375

02615659

2615665

02652646

02654241

02657645

2667837

02670018

02670539

02672175

2673298

02679341

02684169

Work Orders

C92742989

C92822092

C93320212

C93320213

C93399550

C93465583

Miscellaneous

CS-ISC*05, Reactor Water Level - Off Normal Calibration Conditions, Revision 1

General Electric-Hitachi Service Information Letter 678, Susceptibility of Original Directional Control Valve Cap Screw N170P23012C6 to Stress Corrosion Cracking, Revision 0 NRC Bulletin 93-03: Resolution of Issues Related to Reactor Vessel Water Level Instrumentation in BWRs NRC Generic Letter 92-04: Resolution of Issues Related to Reactor Vessel Water Level Instrumentation in BWRs Pursuant to 10

CFR 50.54(F) U2 TS

Section 1R18: Plant Modifications

Procedures

N2-ISP-LDS-Q006, Main Steam Line Tunnel and MSL Lead Enclosure Temperature and Main Steam Line Differential Temperature Instrument Channel Functional Test, Revision 007-00 N2-MFT-323A, Riley Temperature Switch to Yokogawa Panels 2CEC*PNL609/2CEC*PNL611, Revision 00200 N2-OSP-CNT-R003, Containment Isolation Valve Isolation Actuation Surveillance, Revision 01300

Drawings

PID-61A, Piping & Instrumentation Diagram Primary Containment Purge & Standby Gas Treatment, Revision 16

PID-61B, Piping & Instrumentation Diagram Primary Containment Purge & Standby Gas Treatment, Revision 23

PID-61D, Piping & Instrumentation Diagram Primary Containment Purge & Standby Gas Treatment, Revision 00

Issue Reports

02512213

2602387

02607569

02613114

02644004

02656258

02657082

02657445

02657753

02658172

Work Orders

C92545566

Miscellaneous

ECP-14-000837, Engineering Equivalency to Replace Capacitor on 2FWS-PNL10A,B,C for Lockup Circuit, Revision 0

ECP-15-000087, Fukushima Hardened Vents

ECP-15-000434, Replace Riley Temperature Switches with Yokogawa Recorders, Revision 0000

ECP-15-000434-103-01, Design Consideration Summary, Revision 001

ECP-16-000272, Appendix J LLRT Scope Reduction, Revision 0000

ECP-15-000699, Evaluation of Smoke Detection Capability in the NMP2 Relay Room on March 30, 2016 Local Leak Rate Testing Scope Reduction for CSH, CSL, ICS, RDS, RHS, SLS, and WCS Systems Technical Specification 5.5.12, "10

CFR 50 Appendix J Testing Program Plan" Nine Mile Point Generation Station, Unit 2 (Draft R03), True North Consulting,

LLC 5059-2015-077, Hardened Containment Vent System Design 50.59 Screening Form, Revision 0002

Section 1R19: Post-Maintenance Testing

Procedures

N1-MMP-028-105, Control Rod Drive Pump Maintenance

PMP 28 15 (#11) and

PMP 28 17 (#12), Revision 00800 N2-EPM-GEN-635, Operational Checks, Cleaning, and Inspecting Unit 2 UPS's, Revision 00600

N2-IMP-FWS-002, Functional Test and Calibration of 2FWS-LV10A/B and C Limitorque Actuators, Revision 00500 N2-IPM-LRT-@001, Leak Rate Monitor Flow and Electrical Integrity Check, Revision 00300

N2-ISP-MSS-R@002, Main Steam Isolation Valve Leak Rate Test (Reactor Vessel Head Removed), Revision 01000 N2-ISP-MSS-R@003, Main Steam Isolation Valve Leak Rate Test (Inboard Static Head of Water), Revision 00202 N2-MFT-315, Uninterruptible Power Supply 2VBB-UPS3A, Revision 00100

N2-MFT-322, EHC Vertical Pump Modification Test, Revision 00000 N2-MSP-ADS-2Y003, ADS Safety Relief Valve Operability Test, Revision 01 N2-OSP-ADS-R001, ADS Valve and Position Indication Operability Test, Revision 00601

Drawings

C-18011-C Sheet 2, Instrument Air System P&I Diagram, Revision 49

C-23153-C, Turbine Building High Pressure Instrument Piping for Instrument & Service Air Compressors, Revision 1

Issue Reports

02657647

2657697

02674568

Work Orders

C91903530

C92740922

C92741720

C92744486

C92844517

C92984760

C93061396

C93202891

91903530

Miscellaneous

ECP-15-000389, ERVR EHC Upgrades, Revision 0001

GEK-46355E, Hydraulic Power Unit

PCR-16-01399, Uninterruptible Power Supply 2VBB-UPS 3A, Revision 100

U2 UFSAR

Section 1R20: Refueling and Other Outage Activities

Procedures

LS-AA-112, Fatigue Management and Work Hour Limits Revision 12

N2-FHP-003, Refueling Manual, Revision 01500 N2-FHP-13.3, Core Shuffle, Revision 00700 N2-OP-101C, Plant Shutdown, Revision 02700N2-OP-101D, Power Changes, Revision 02200

Section 1R22: Surveillance Testing

Procedures

CY-AA-130-3010-F-03, Dose Equivalent Iodine Determination, Revision 3

CY-AB-120-1000, BWR Strategic Water Chemistry Plan, Revision 12 N1-CSP-D100, Reactor Coolant Chemistry, Revision 1300 N1-ST-Q8A, Liquid Poison Pump 11 and Check Valve Operability Test, Revision 01

N2-ESP-ENS-Q731, Quarterly Channel Functional Test of LPCS/LPCI Pumps A, B, and C (normal and emergency power) Auto Start Time Delay Relays, Revision 00700

N2-ISP-MSS-R002, Main Steam Isolation Valve Leak Rate Test (Reactor Vessel Head Removed), Revision 00902 N2-ISP-MSS-R@003, Main Steam Isolation Valve Leak Rate Test (Inboard Static Head of Water), Revision 00202 N2-OSP-EGS-R004, Operating Cycle Diesel Generator Simulated Loss of Offsite Power with ECCS Division 1 & 2, Revision 01500 N2-OSP-ICS-R002, RCIC System Flow Test, Revision 00400

N2-OSP-MSS-CS001, Main Steam Isolation Valve Operability, Revision 00700 N2-OSP-RPS-R301, RPS Logic System Functional Test, Revision 00500 N2-OSP-SFC-Q001, Spent Fuel Pool Cooling Cleanup Pump and Valve Operability Test, Revision 00600 N2-OSP-SLS-Q001, Standby Liquid Control Pump, Check Valve, Relief Valve Operability Test and ASME XI Pressure Test, Revision 01600

Issue Reports

02659867

02665932

2684642

Work Orders

C92740922

C93084076

C93093043

Section 2RS1: Radiological Hazard Assessment and Exposure Controls

Procedures

RP-AA-19, High Radiation Area Program Description, Revision 2

RP-AA-100, Conduct of Radiation Protection Operations, Revision 3

RP-AA-300, Radiological Survey Program, Revision 13

RP-AA-376, Radiological Postings, Labeling, and Markings, Revision 8

RP-AA-460, Controls for High and Locked High Radiation Areas, Revision 28

RP-AA-460-001, Controls for Very High Radiation Areas, Revision 6

RP-AA-460-002, Additional High Radiation Exposure Control, Revision 3

RP-AA-800, Control, Inventory, and Leak Testing of Radioactive Sources, Revision 7

RP-AA-1008, Unescorted Access to and Conduct in Radiologically Controlled Areas, Revision 5

Issue Reports

02652134

2654590

2638570

Miscellaneous

Declaration of Intent to Become Pregnant, March 11, 2016

DW 249' Survey, April 18, 2016

Human Performance Review Board (HU-AA-101-1002), April 11, 2016 MSIV 7C Breach Survey, April 19, 2016 MSIV 7C Post Decontamination Survey, April 21, 2016

MSIV 6D Breach Survey, April 20, 2016 Radwaste Building 265' Survey, March 9, 2016

Unit 1 Reactor Building Weekly Survey, April 19, 2016 and March 16, 2016

Unit 1 Turbine Building Weekly Survey, April 20, 2016 and March 18, 2016

Section 2RS2: Occupational

ALARA Planning and Controls

Procedures

RP-AA-400, ALARA Program, Revision 13

RP-AA-400-1006, Outage Exposure Estimating and Tracking, Revision 6

RP-AA-403, Administration of the Radiation Work Permit Program, Revision 8

Issue Reports

02641361

Miscellaneous

ALARA 50% Work in Progress Review, 2016-12 ALARA Plan 2016-2-012, N2R15 CRD Activities

ALARA Plan 2016-2-014 N2R15 Drywell Permanent Shielding Modification

ALARA Plan 2016-2-018 N2R15 DW ISI Exams and Support Activities

ALARA Plan 2016-2-020 N2R15 Drywell Snubber Testing and Inspections

Section 2RS3: In-Plant Airborne Radioactivity Control and Mitigation

Procedures

RP-AA-825-1020, Operation and Use of Air Line Supplied Respirators Issue Reports

02642282

Miscellaneous

NMPNPS Approval to Use Mururoa Air Supplied Suits, January 19, 2006

Honeywell Monitor T25487 Grade D Air Sample Analysis

Section 4OA1: Performance Indicator Verification

Miscellaneous

NEI-99-02, Regulatory Assessment Performance Indicator Guideline, Revision 7

Section 4OA2: Problem Identification and Resolution

Procedures

LEP-DG-105, Maintenance and Adjustment of Woodward U 8 Governor Shutdown Solenoids, Revision 13 N1-ST-M4B, Emergency Diesel Generator 103 and

PB 103 Operability Test, Revision 01700

N1-EPM-DGE-252, Emergency Diesel Generator Inspection (DG102 and DG103), Revision 01200 N1-RCSP-GEN-334, Operating Cycle Calibration for Loss and Degraded Voltage Relays on Emergency Switchgear, Revision 00300

PI-AA-115, Operating Experience Program, Revision 1

PI-AA-125, Corrective Action Program, Revision 4

PI-AA-120, Issue Identification and Screening Process, Revision 6

PI-AA-125-1003, Apparent Cause Evaluation Manual, Revision 3

Drawings

C-19410-C Sheet 10, Elementary Wiring Diagram 4.16KV Emergency Power Boards & Diesel Generators (#102 & #103 Control Circuits), Revision 44.01

Issue Reports

00828185

02613086

02621942

2622472

2631968

02640675

02648466

02664004

2668118

2670006

02671608

02674427

02675958

2678327

2682226

Work Orders

C92393825

C92653166

Miscellaneous

ECP-16-000443, Technical Evaluation to Summarize EDG Operability under

IR 02640675, Revision 0

DRIFT-27N-59N, Drift Analysis -

ABB 27N and 59N Series Undervoltage and Overvoltage Relays, Revision 0 Nine Mile Point Nuclear Station Unit 1 GNF2

ECCS-LOCA Evaluation with Sustained Under-Voltage Condition, January 2013 Nine Mile Point Nuclear Station Unit 1 GNF2

ECCS-LOCA Evaluation, January 2014

NOSA-NMP-16-01, Maintenance Program Audit Report

SDBD-804, Emergency Diesel Generator System, Revision 11

System Health Report, NMP Unit 2, EDG, Q1-2016 Unit 1 Emergency Diesel Generator System Health Report 1st Quarter 2016 Woodward Product Manual 03013, Shutdown Solenoid for UG Governor, Revision T

Section 4OA3: Follow-up of Events and Notices of Enforcement Discretion

Procedures

HU-AA-102, Technical Human Performance Practices, Revision 007

N2-ISP-ISC-Q005, Quarterly Functional Test of Reactor Vessel Level 2 and Level 1 Instrument Channels, Revision 0600

PI-AA-120, Issue Identification and Screening Process, Revision 6

WC-AA-106, Work Screening and Processing, Revision 16

Issue Reports

01700638

01984948

02006657

02617343

02652511

Work Orders

C91856157

C92671987

C92681851

C93247505

Miscellaneous

ECP-14-000514, Reactor Water Cleanup Isolation Valve Control Circuit, Revision 0 N1-NFPA-003, Unfused Control Circuit

IV-33-04, Revision 0

NEI 99-02, Regulatory Assessment Performance Indicator Guideline, Revision 7

Nine Mile Point, Unit 1 Facility Operating License

LIST OF ACRONYMS

CFR Title 10 of the Code of Federal Regulations

AC alternating current

ALARA as low as reasonably achievable

ASME American Society of Mechanical Engineers

CAP corrective action program

CR condition report

DC direct current

ECCS emergency core cooling system

ECP engineering change package

EDG emergency diesel generator

EMD electromotive division

FA fire area

GL generic letter

HPCS high-pressure core spray

HURB human performance review board

HX heat exchanger

IMC Inspection Manual Chapter

ISI inservice inspection

kV kilovolt

LAR license amendment request

LCO limiting condition for operation

NCV non-cited violation

NDE non-destructive examination

NEI Nuclear Energy Institute

NMPNS Nine Mile Point Nuclear Station,

LLC [[]]

NRC Nuclear Regulatory Commission,

U.S. [[]]

OOS out of service

OWA operator workaround P&

ID piping and instrument drawing

psig per square inch gauge

PI performance indicator

PMT post maintenance test

RCS reactor coolant system

RCIC reactor core isolation cooling

RHR residual heat removal

RPV reactor pressure vessel

RRMG reactor recirculation motor generator

RT radiographic testing

SSC structure, system, and component

SSFF safety system functional failure

TS technical specification

UFSAR updated final safety analysis report

UPS uninterruptible power supply

UT ultrasonic testing

VT visual testing

WO work order

@@ Line 2: / Line 2: @@
 | number = ML16221A101
 | issue date = 08/08/2016
-| title = Nine Mile Point Nuclear Station - Integrated Inspection Report 05000220/2016002 And 05000410/2016002 And Exercise Of Enforcement Discretion
+| title = Nine Mile Point Nuclear Station - Integrated Inspection Report 05000220/2016002 and 05000410/2016002 and Exercise of Enforcement Discretion
 | author name = Scott M L
 | author affiliation = NRC/RGN-I/DRP

IR 05000220/2016002: Difference between revisions

Revision as of 23:07, 1 March 2018

Text

Dear Mr. Hanson:

Enclosure:

w/Attachment:

SUMMARY

Cornerstone: Initiating Events

Cornerstone: Mitigating Systems

Cornerstone: Occupational Radiation Safety

REPORT DETAILS

REACTOR SAFETY

1R01 Adverse Weather Protection

.1 Readiness for Seasonal Extreme Weather Conditions

a. Inspection Scope

b. Findings

.2 Summer Readiness of Offsite and Alternate Alternating Current (AC) Power Systems

a. Inspection Scope

b. Findings

.3 External Flooding (2 samples)

a. Inspection Scope

b. Findings

Introduction.

Description.

Analysis.

Enforcement.

.1 Partial System Walkdowns

a. Inspection Scope

b. Findings

.2 Full System Walkdown

a. Inspection Scope

b. Findings

1R05 Fire Protection

.1 Resident Inspector Quarterly Walkdowns

a. Inspection Scope

b. Findings

.2 Fire Protection - Drill Observation

a. Inspection Scope

b. Findings

1R06 Flood Protection Measures

a. Inspection Scope

b. Findings

1R07 Heat Sink Performance

a. Inspection Scope

b. Findings

1R08 Inservice Inspection

a. Inspection Scope

b. Findings

1R11 Licensed Operator Requalification Program and Licensed Operator Performance

.1 Quarterly Review of Licensed Operator Requalification Testing and Training (2 samples)

a. Inspection Scope

b. Findings

.2 Quarterly Review of Licensed Operator Performance in the Main Control Room

a. Inspection Scope

b. Findings

1R12 Maintenance Effectiveness

a. Inspection Scope

b. Findings

1R13 Maintenance Risk Assessments and Emergent Work Control

a. Inspection Scope

b. Findings

1R15 Operability Determinations and Functionality Assessments

a. Inspection Scope

b. Findings

Introduction.

Description.

Analysis.

Enforcement.

1R18 Plant Modifications

.1 Temporary Modifications (1 sample)

a. Inspection Scope

b. Findings

.2 Permanent Modifications (4 samples)

a. Inspection Scope

b. Findings

1R19 Post-Maintenance Testing

a. Inspection Scope

b. Findings

1R20 Refueling and Other Outage Activities

a. Inspection Scope