Text

NRC Integrated Inspection Report 05000445/2016002 and 05000446/2016002
	ML16224A535
Person / Time
Site:	Comanche Peak
Issue date:	08/11/2016
From:	Jeremy Groom; Division of Reactor Safety IV
To:	Peters K; Luminant Generation Co
	Jeremy Groom
References
	IR 2016002
	Download: ML16224A535 (40)
	v • d • e

UNITED STATES ust 11, 2016

SUBJECT:

COMANCHE PEAK NUCLEAR POWER PLANT - NRC INTEGRATED INSPECTION REPORT 05000445/2016002 and 05000446/2016002

Dear Mr. Peters:

On June 30, 2016, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at your Comanche Peak Nuclear Power Plant, Units 1 and 2. On July 7, 2016, the NRC inspectors discussed the results of this inspection with Mr. Tom McCool, Site Vice President, and other members of your staff. Inspectors documented the results of this inspection in the enclosed inspection report.

NRC inspectors documented two findings of very low safety significance (Green) in this report.

Both of these findings involved violations of NRC requirements. Further, inspectors documented two licensee-identified violations which were determined to be of very low safety significance in this report. The NRC is treating these violations as non-cited violations (NCVs)

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

If you contest the violations or significance of these NCVs, you should provide a response within 30 days of the date of this inspection report, with the basis for your denial, to the U.S. Nuclear Regulatory Commission, ATTN: Document Control Desk, Washington, DC 20555-0001; with copies to the Regional Administrator, Region IV; the Director, Office of Enforcement, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001; and the NRC resident inspector at the Comanche Peak Nuclear Power Plant, Units 1 and 2.

If you disagree with a cross-cutting aspect assignment in this report, you should provide a response within 30 days of the date of this inspection report, with the basis for your disagreement, to the Regional Administrator, Region IV; and the NRC resident inspector at the Comanche Peak Nuclear Power Plant, Units 1 and 2.

In accordance with Title 10 of the Code of Federal Regulations (10 CFR) 2.390, Public Inspections, Exemptions, Requests for Withholding, a copy of this letter, its enclosure, and your response (if any) will be available electronically for public inspection in the NRCs Public Document Room or from the Publicly Available Records (PARS) component of the NRC's Agencywide Documents Access and Management System (ADAMS). ADAMS is accessible from the NRC Web site at http://www.nrc.gov/reading-rm/adams.html (the Public Electronic Reading Room).

Sincerely,

/RA/

Jeremy R. Groom, Branch Chief Project Branch A Division of Reactor Projects Docket Nos. 50-445 and 50-446 License Nos. NPF-87 and NPF-89 Enclosure:

Inspection Report 05000445/2016002 and 050446/2016002 w/ Attachment: Supplemental Information

ML16224A535 SUNSI Review ADAMS Non-Sensitive Publicly Available By: JGroom Yes No Sensitive Non-Publicly Available OFFICE SRI:DRP/A SPE:DRP/A BC:EB1 BC:EB2 BC:OB BC:PSB2 TL-IPAT NAME JJosey RAlexander TFarnholtz GWerner VGaddy HGepford THipschman SIGNATURE /RA/ /RA/ /RA/ /RA/ /RA/ /RA/ /RA/ HAF for DATE 7/24/2016 7/24/2016 7/25/2016 7/25/2016 7/25/2016 7/25/2016 7/25/2016 OFFICE BC:DRP/A NAME JGroom SIGNATURE /RA/

DATE 8/10/2016

Letter to Ken Peters from Jeremy Groom dated August 11, 2016 SUBJECT: COMANCHE PEAK NUCLEAR POWER PLANT-NRC INTEGRATED INSPECTION REPORT 05000445/2016002 and 05000446/2016002 DISTRIBUTION:

Regional Administrator (Kriss.Kennedy@nrc.gov)

Deputy Regional Administrator (Scott.Morris@nrc.gov)

DRP Director (Troy.Pruett@nrc.gov)

DRP Deputy Director (Ryan.Lantz@nrc.gov)

DRS Director (Anton.Vegel@nrc.gov)

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

Senior Resident Inspector (Jeffrey.Josey@nrc.gov)

Resident Inspector (Rayomand.Kumana@nrc.gov)

Administrative Assistant (Rhonda.Smith@nrc.gov)

Branch Chief, DRP/A (Jeremy.Groom@nrc.gov)

Senior Project Engineer, DRP/A (Ryan.Alexander@nrc.gov)

Project Engineer, DRP/A (Thomas.Sullivan@nrc.gov)

Project Engineer, DRP/A (Mathew.Kirk@nrc.gov)

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

Project Manager (Margaret.Watford@nrc.gov)

Team Leader, DRS/IPAT (Thomas.Hipschman@nrc.gov)

RITS Coordinator (Marisa.Herrera@nrc.gov)

ACES (R4Enforcement.Resource@nrc.gov)

Regional Counsel (Karla.Fuller@nrc.gov)

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

Congressional Affairs Officer (Jenny.Weil@nrc.gov)

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

RIV/ETA: OEDO (Jeremy.Bowen@nrc.gov)

ROPreports Electronic Distribution for Comanche Peak Nuclear Power Plant

U.S. NUCLEAR REGULATORY COMMISSION

REGION IV

Docket: 05000445, 05000446 License: NPF-87, NPF-89 Report: 05000445/2016002 and 05000446/2016002 Licensee: Luminant Generation Company, LLC Facility: Comanche Peak Nuclear Power Plant, Units 1 and 2 Location: 6322 N. FM-56, Glen Rose, Texas Dates: April 1 through June 30, 2016 Inspectors: J. Josey, Senior Resident Inspector S. Hedger, Acting Senior Resident Inspector R. Kumana, Resident Inspector W. Cullum, Reactor Inspector J. Drake, Senior Reactor Inspector P. Hernandez, Health Physicist M. Phalen, Senior Health Physicist Approved Jeremy R. Groom By: Chief, Project Branch A Division of Reactor Projects Enclosure

SUMMARY

IR 05000445/2016002 and 05000446/2016002; 04/01/2016 - 06/30/2016; Comanche Peak

NPP, Units 1 and 2; Surveillance Testing, Radiological Hazard Assessment and Exposure Controls The inspection activities described in this report were performed between April 1, 2016, through June 30, 2016, by the resident inspectors at the Comanche Peak Nuclear Power Plant and inspectors from the NRCs Region IV office. Two findings of very low safety significance (Green) are documented in this report. All of these findings involved a violation of NRC requirements. Additionally, NRC inspectors documented in this report two licensee-identified violations of very low safety significance. The significance of inspection findings is indicated by their color (Green, White, Yellow, or Red), which is determined using Inspection Manual Chapter 0609, Significance Determination Process. Their cross-cutting aspects are determined using Inspection Manual Chapter 0310, Aspects within the Cross-Cutting Areas.

Violations of NRC requirements are dispositioned in accordance with the NRC Enforcement Policy. The NRCs program for overseeing the safe operation of commercial nuclear power reactors is described in NUREG-1649, Reactor Oversight Process.

Cornerstone: Mitigating Systems

Green.

The inspectors identified a non-cited violation of 10 CFR 50, Appendix B,

Criterion XVI, Corrective Action, associated with the licensees failure to correct a condition adverse to quality in safety-related equipment. Specifically, following an in-service testing failure of auxiliary feedwater check valve 2FW-091 in November 2012, the licensee performed an operability evaluation of the auxiliary feedwater system. However, the inspectors identified that the licensee failed to take corrective action to address the condition adverse to quality that resulted in the valve failing to seat properly. Consequently, the same valve failed a subsequent inservice test in November 2015. Following discovery of this issue, the licensee performed an operability determination that established a reasonable expectation of operability pending implementation of corrective actions. The licensee entered this issue into corrective action program as CR-2015-10961.

The licensees failure to correct a condition adverse to quality was a performance deficiency.

The performance deficiency was more than minor, and therefore a finding, because it was associated with the equipment performance attribute of the Mitigating Systems Cornerstone and affected the cornerstone objective to ensure availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the licensee failed to correct auxiliary feedwater check valve 2FW-0191 failure to seat in November 2012 resulting in an additional failure in November 2015. Using Inspection Manual Chapter (IMC) 0609, Appendix A, The Significance Determination Process (SDP)for Findings At-Power, dated June 19, 2012, inspectors determined that this finding was of very low safety significance (Green) because the finding (1) was not a deficiency affecting the design and qualification of a mitigating structure, system, or component, and did not result in a loss of operability or functionality, (2) did not represent a loss of system and/or function, (3) did not represent an actual loss of function of at least a single train for longer than its allowed outage time, or two separate safety systems out-of-service for longer than their technical specification allowed outage time, and (4) did not represent an actual loss of function of one or more non-technical specification trains of equipment designated as high safety-significant for greater than 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months in accordance with the licensees maintenance rule program. The finding has a problem identification and resolution cross-cutting aspect associated with evaluation, in that, the licensee failed to thoroughly evaluate issues to ensure that resolutions address extent of conditions. Specifically, the licensee failed to appropriately classify the issue of the check valve not seating and recognize this as a degraded condition [P.2]. (Section 1R22)

Cornerstone: Occupational Radiation Safety

Green.

The inspectors reviewed a self-revealed non-cited violation of Technical Specification 5.7.1.e associated with the licensee allowing a worker access into the 2-077-B penetration valve room, a high radiation area, without an adequate knowledge of the radiological conditions. Specifically, the licensee briefed the worker on the conditions with outdated radiation survey information even though the 2-077-B penetration valve room was subject to changing radiological conditions. As a result, an individual entered areas with general area dose rates of 210 mrem per hour rather than the briefed dose rates of less than 50 mrem per hour. This issue was entered into the licensees corrective action program as Condition Report CR-2015-010211. Corrective actions included performing follow-up radiation surveys and implementing improvements to the high radiation area access control program.

The inspectors determined that allowing a worker access into a high radiation without an adequate knowledge of the radiological conditions was a performance deficiency. The performance deficiency was more than minor, and therefore a finding, because it affected the program and process attribute of the Occupational Radiation Safety cornerstone and adversely affected the cornerstone objective of ensuring adequate protection of worker health and safety from exposure to radiation. Specifically, entry into a high radiation area without adequate knowledge of the radiological conditions placed the individual at risk for unnecessary exposure. The finding was assessed using Inspection Manual Chapter 0609,

Appendix C, Occupational Radiation Safety Significance Determination Process, issued August 19, 2008, and was determined to be of very low safety significance (Green) because the performance deficiency was not an ALARA planning issue, there was not an overexposure nor substantial potential for an overexposure, and the licensees ability to assess dose was not compromised. The finding has a human performance cross-cutting aspect associated with work management, because the organization failed to implement a process of planning, controlling, and executing work activities such that nuclear safety was the overriding priority [H.5]. (Section 2RS1)

Licensee-Identified Violations

Violations of very low safety significance that were identified by the licensee have been reviewed by the inspectors. Corrective actions taken or planned by the licensee have been entered into the licensees corrective action program. These violations and associated corrective action tracking numbers are listed in Section 4OA7 of this report.

PLANT STATUS

Unit 1 began the inspection period at approximately 100 percent power. On April 26, 2016, the unit commenced coastdown to a planned refueling outage. The 1RF18 outage began with the normal reactor shutdown on May 1, 2016. The unit remained in an outage until May 31, 2016, when the generator breaker was closed and the unit synchronized to the grid. Unit 1 ramped to approximately 100 percent power on June 3, 2016, and operated at that power for the remainder of the inspection period.

Unit 2 began the inspection period at approximately 100 percent power and operated at that power level for the entire inspection period.

REPORT DETAILS

REACTOR SAFETY

Cornerstones: Initiating Events, Mitigating Systems, and Barrier Integrity

1R01 Adverse Weather Protection

Main article: IP 71111.01

.1 Readiness for Impending Adverse Weather Conditions

a. Inspection Scope

On April 21, 2016, the inspectors completed an inspection of the stations readiness for impending adverse weather conditions involving heavy rains. The inspectors reviewed plant design features, the licensees procedures to respond to Squaw Creek Reservoir level rising above 776 feet, and the licensees implementation of these procedures. The inspectors evaluated operator staffing and accessibility of controls and indications for those systems required to control the plant.

On April 26, 2016, the inspectors completed an additional inspection of the stations readiness for impending adverse weather conditions involving severe thunderstorms and high winds. The inspectors reviewed plant design features, the licensees procedures to respond to severe thunderstorms, large hail, and tornados, and the licensees planned implementation of these procedures. The inspectors evaluated operator staffing and accessibility of controls and indications for those systems required to control the plant.

These activities constituted two samples of readiness for impending adverse weather conditions, as defined in Inspection Procedure 71111.01.

Findings No findings were identified.

1R04 Equipment Alignment

Main article: IP 71111.04

.1 Partial Walkdown

a. Inspection Scope

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

March 29, 2016, Unit 1, atmospheric relief valves for steam generator 1-01, 1-02, and 1-03

April 21, 2016, Unit 2, diesel generator 2-02 jacket water piping system The inspectors reviewed the licensees procedures and system design information to determine the correct lineup for the systems. They visually verified that critical portions of the systems or trains were correctly aligned for the existing plant configuration.

These activities constituted two partial system walk-down samples as defined in Inspection Procedure 71111.04.

b. Findings

No findings were identified.

1R05 Fire Protection

Main article: IP 71111.05

.1 Quarterly Inspection

a. Inspection Scope

The inspectors evaluated the licensees fire protection program for operational status and material condition. The inspectors focused their inspection on four plant areas important to safety:

April 11, 2016, Unit 2, Fire area EM, cable spreading room

April 12, 2016, Unit 1, Fire area EN, cable spreading room

April 19, 2016, Units 1 and 2, Fire Zone ER150, Train A uninterruptible power supply HVAC room

April 19, 2016, Unit 1 Fire area AA39, ventilation filter room For each area, the inspectors evaluated the fire plan against defined hazards and defense-in-depth features in the licensees fire protection program. The inspectors evaluated control of transient combustibles and ignition sources, fire detection and suppression systems, manual firefighting equipment and capability, passive fire protection features, and compensatory measures for degraded conditions.

These activities constituted four quarterly inspection samples, as defined in Inspection Procedure 71111.05.

b. Findings

No findings were identified.

1R07 Heat Sink Performance

Main article: IP 71111.07

a. Inspection Scope

On June 27, 2016, the inspectors completed an inspection of the readiness and availability of risk-significant heat exchangers. The inspectors verified the licensee used the industry standard periodic maintenance method outlined in EPRI NP-7552 for the diesel generator 1-02 jacket water heat exchanger, and observed the licensees inspection of the diesel generator 1-02 jacket water heat exchanger and the material condition of the heat exchanger internals. Additionally, the inspectors walked down the diesel generator 1-02 jacket water heat exchanger to observe its performance and material condition.

These activities constitute completion of one heat sink performance annual review sample, as defined in Inspection Procedure 71111.07.

b. Findings

No findings were identified.

1R08 Inservice Inspection Activities

Main article: IP 71111.08

The activities described in Subsections 1 through 4 below constitute completion of one inservice inspection sample, as defined in Inspection Procedure 71111.08.

.1 Non-Destructive Examination Activities and Welding Activities

a. Inspection Scope

The inspectors directly observed the following nondestructive examinations:

SYSTEM WELD IDENTIFICATION EXAMINATION TYPE Main Steam TBX-2-2100-2B-W28 Ultrasonic Test Main Steam TBX-2-2100-2B-W31 Ultrasonic Test Residual Heat Removal TBX-1-4101-W5 Ultrasonic Test Residual Heat Removal TBX-1-4101-W6 Ultrasonic Test Safety Injection TBX-2-2533-H-6 Visual Examination Safety Injection TBX-2-2533-H-8 Visual Examination Safety Injection TBX-2-2533-H-9 Visual Examination Safety Injection TBX-2-2533-H-11 Visual Examination The inspectors reviewed records for the following nondestructive examinations:

SYSTEM WELD IDENTIFICATION EXAMINATION TYPE Containment Spray TBX-2-1180-2-2WS Penetrant Test Containment Spray TBX-2-1180-2-2-1 Penetrant Test Containment Spray TBX-2-1180-2-2-2 Penetrant Test Containment Spray TBX-2-1180-2-2-3 Penetrant Test Containment Spray TBX-2-1180-2-2-4 Penetrant Test Reactor Coolant System TBX-1-1300A-SUP Visual Examination Main Steam TBX-2-2100-MR1 Visual Examination Feedwater TBX-2-2101-H7 Visual Examination Service Water SW-1-AB-010-H1 Visual Examination Containment Spray TBX-2-1180-2-2 Ultrasonic Test Residual Heat Removal TBX-2-2532-5 Ultrasonic Test Residual Heat Removal TBX-2-2532-34 Ultrasonic Test Residual Heat Removal TBX-2-2532-35 Ultrasonic Test Residual Heat Removal TBX-2-253236 Ultrasonic Test During the review and observation of each examination, the inspectors observed whether activities were performed in accordance with the ASME Code requirements and applicable procedures. The inspectors reviewed two indications that were previously examined and observed that the licensee evaluated and accepted the indications in accordance with the ASME Code. The inspectors also reviewed the qualifications of all nondestructive examination technicians performing the inspections and determined that they were current.

There were no welding activities performed while the inspector was on site.

Main article: IP 71111.11

.1 Review of Licensed Operator Requalification

a. Inspection Scope

On April 4, 2016, the inspectors observed a portion of an annual simulator requalification test for licensed operators. The inspectors assessed the performance of the operators and the evaluators critique of their performance. The inspectors also assessed the modeling and performance of the simulator during the requalification activities.

These activities constitute completion of one quarterly licensed operator requalification program sample, as defined in Inspection Procedure 71111.11.

b. Findings

No findings were identified.

.2 Review of Licensed Operator Performance

a. Inspection Scope

Inspectors observed the performance of on-shift licensed operators in the plants main control room. At the time of the observations, the plant was in a period of heightened activity or risk due to outage activities. The inspectors observed the operators performance of the following activities:

May 5, 2016, Unit 1, reactor cavity fill with one residual heat removal train

May 25, 2016, Unit 1, reduced inventory operations

May 30, 2016, Unit 1, reactor startup activities In addition, the inspectors assessed the operators adherence to plant procedures, including conduct of operations procedure and other operations department policies.

These activities constitute completion of one quarterly licensed operator performance sample, as defined in Inspection Procedure 71111.11.

b. Findings

No findings were identified.

1R12 Maintenance Effectiveness

Main article: IP 71111.12

a. Inspection Scope

The inspectors reviewed one instance of degraded performance or condition of safety-related structures, systems, and components (SSCs):

June 30, 2016, Unit 2 safety injection system, high unavailability due to unplanned maintenance The inspectors reviewed the extent of condition of possible common cause SSC failures and evaluated the adequacy of the licensees corrective actions. The inspectors reviewed the licensees work practices to evaluate whether these may have played a role in the degradation of the SSCs. The inspectors assessed the licensees characterization of the degradation in accordance with 10 CFR 50.65 (the Maintenance Rule), and verified that the licensee was appropriately tracking degraded performance and conditions in accordance with the Maintenance Rule.

These activities constituted completion of one maintenance effectiveness sample, as defined in Inspection Procedure 71111.12.

b. Findings

No findings were identified.

1R13 Maintenance Risk Assessments and Emergent Work Control

Main article: IP 71111.13

a. Inspection Scope

The inspectors reviewed four risk assessments performed by the licensee prior to changes in plant configuration and the risk management actions taken by the licensee in response to elevated risk:

March 8, 2016, Unit 1 and 2, elevated risk for adverse weather

April 27, 2016, Unit 1, refueling outage 1RF18 defense in depth plan

May 9, 2016, Unit 1, decay heat removal configuration during orange risk window

June 22, 2016, Unit 1, barrier impairment for opening the Unit 1 service water tunnel door The inspectors verified that these risk assessments were performed timely and in accordance with the requirements of 10 CFR 50.65 (the Maintenance Rule) and plant procedures. The inspectors reviewed the accuracy and completeness of the licensees risk assessments and verified that the licensee implemented appropriate risk management actions based on the result of the assessments.

Additionally, on April 6, 2016, the inspectors also observed portions of an emergent work activity for diesel generator 2-02 that had the potential to affect the functional capability of a mitigating system.

The inspectors verified that the licensee appropriately developed and followed a work plan for these activities. The inspectors verified that the licensee took precautions to minimize the impact of the work activities on SSCs.

These activities constitute completion of five maintenance risk assessments and emergent work control inspection samples, as defined in Inspection Procedure 71111.13.

b. Findings

No findings were identified.

1R15 Operability Determinations and Functionality Assessments (71111.1)

a. Inspection Scope

The inspectors reviewed seven operability determinations and functionality assessments that the licensee performed for degraded or nonconforming SSCs:

March 10, 2016, CR-2016-002284, operability determination of service water pump 2-02 with wetted termination box

April 13, 2016, CR-2016-003303, operability determination of diesel generator 1-01 with hot web deflection results out of tolerance

April 19, 2016, CR-2016-003516, functionality assessment of compensatory measures for main steam line 1-04 radiation monitor out of service

May 1, 2016, CR-2016-003907, operability determination of reactor trip breakers following high trip response times

May 26, 2016, CR-2016-005248, operability determination for turbine driven auxiliary feedwater pumps due to flash tank vents tornado missile susceptibility

June 21, 2016, CR-2016-004865, operability determination of service water piping for Unit 1 and 2 rust indications

June 22, 2016, CR-2016-004792, operability determination of source range channel N31 The inspectors reviewed the timeliness and technical adequacy of the licensees evaluations. Where the licensee determined the degraded SSC to be operable or functional, the inspectors verified that the licensees compensatory measures were appropriate to provide reasonable assurance of operability or functionality. The inspectors verified that the licensee had considered the effect of other degraded conditions on the operability or functionality of the degraded SSC.

The inspectors reviewed operator actions taken or planned to compensate for degraded or nonconforming conditions. Specifically, the inspectors reviewed the actions of operators with the pressurizer power operated relief valves in manual following a failure of 1-PT-405, as documented in CR-2016-003740. The inspectors verified that the licensee effectively managed these operator workarounds to prevent adverse effects on the function of mitigating systems and to minimize their impact on the operators ability to implement abnormal and emergency operating procedures.

These activities constitute completion of eight operability and functionality review samples, which included one operator work-around sample, as defined in Inspection Procedure 71111.15.

b. Findings

No findings were identified.

1R19 Post-Maintenance Testing

Main article: IP 71111.19

a. Inspection Scope

The inspectors reviewed three post-maintenance testing activities that affected risk-significant SSCs:

April 7, 2016, spent fuel pool cooling pump X-02 following seal replacement

April 7, 2016, diesel generator 2-02 following emergent maintenance

May 28, 2016, steam generator blowdown high energy line break isolation valve following elastomer and packing replacement The inspectors reviewed licensing- and design-basis documents for the SSCs and the maintenance and post-maintenance test procedures. The inspectors observed the performance of the post-maintenance tests to verify that the licensee performed the tests in accordance with approved procedures, satisfied the established acceptance criteria, and restored the operability of the affected SSCs.

These activities constitute completion of three post-maintenance testing inspection samples, as defined in Inspection Procedure 71111.19.

b. Findings

No findings were identified.

1R20 Refueling and Other Outage Activities

Main article: IP 71111.20

a. Inspection Scope

During the stations refueling outage that concluded on May 31, 2016, the inspectors evaluated the licensees outage activities. The inspectors verified that the licensee considered risk in developing and implementing the outage plan, appropriately managed personnel fatigue, and developed mitigation strategies for losses of key safety functions.

This verification included the following:

Review of the licensees outage plan prior to the outage

Review and verification of the licensees fatigue management activities

Monitoring of shut-down and cool-down activities

Verification that the licensee maintained defense-in-depth during outage activities

Observation and review of reduced-inventory and mid-loop activities

Observation and review of fuel handling activities

Monitoring of heat-up and startup activities These activities constitute completion of one refueling outage sample as defined in Inspection Procedure 71111.20.

b. Findings

No findings were identified.

1R22 Surveillance Testing

Main article: IP 71111.22

a. Inspection Scope

The inspectors observed five risk-significant surveillance tests and reviewed test results to verify that these tests adequately demonstrated that the SSCs were capable of performing their safety functions:

In-service tests:

June 13, 2016, Unit 2, steam generator spilt flow bypass check valve 2FW-0191 Containment isolation valve surveillance tests:

May 16, 2016, Unit 1, steam generator 1-04 blowdown isolation valve Other surveillance tests:

May 17, 2016, Unit 2, reactor coolant system iodine activity

April 27, 2016, Unit 1, diesel generator 1-02

May 5, 2016, Unit 1, train B integrated test sequence The inspectors verified that these tests met technical specification requirements, that the licensee performed the tests in accordance with their procedures, and that the results of the test satisfied appropriate acceptance criteria. The inspectors verified that the licensee restored the operability of the affected SSCs following testing.

These activities constitute completion of five surveillance testing inspection samples, as defined in Inspection Procedure 71111.22.

b. Findings

Introduction.

The inspectors identified a Green non-cited violation of 10 CFR 50, Appendix B, Criterion XVI, Corrective Action, associated with the licensees failure to correct a condition adverse to quality in safety related equipment. Specifically, following an in-service testing failure of auxiliary feedwater check valve 2FW-0191 in November 2012, the licensee performed an operability evaluation of the auxiliary feedwater system. However, the inspectors identified that the licensee failed to take corrective action to address the condition adverse to quality that resulted in the valve failing to seat properly. Consequently, the same valve failed a subsequent inservice test in November 2015.

Description.

In November 2012, the licensee conducted in-service testing on auxiliary feedwater check valve 2FW-0191, a steam generator split flow bypass check valve.

During the test, check valve 2FW-0191 failed to meet the sites acceptance criteria that allowed for a slight pressure increase following initiation of system flow as long as the pressure increase is less than or equal to 2 percent of the test gauges full scale and steady. During the test, upstream pressure did not remain steady and continued to rise for the duration of the test, indicating that the check valve did not seat. The licensee initiated Condition Report CR-2012-011563 to document the failed test. The licensee performed an operability evaluation for the auxiliary feedwater system but took no other actions to correct the issue associated with the failure of the valve to seat completely.

On November 11, 2015, the licensee conducted the same in-service testing on auxiliary feedwater check valve 2FW-0191. During this test, the licensee observed no differential pressure across check valve 2FW-0191 indicating the valve failed to seat. The licensee stopped the test and initiated Condition Report CR-2015-10961 to document the test failure. The licensee then mechanically agitated the check valve and performed the inservice test again. During the subsequent testing the licensee noted a differential pressure of 470 psid. The licensee documented an operability evaluation for the auxiliary feedwater system to demonstrate a reasonable expectation of operability with the check valve in a degraded condition.

The inspectors determined that the licensee failed to correct a condition adverse to quality. Specifically, inspectors concluded that the use of an operability determination in November 2012 justified continued operation with a degraded condition involving the failure of auxiliary feedwater check valve 2FW-0191 to seat. However, an operability determination is separate from corrective action to restore full qualification.

Consequently, because the licensee only evaluated the operability of the auxiliary feedwater system but took no action to address the degraded condition, the same check valve failed a subsequent inservice test in November 2015. Additionally, the inspectors determined that mechanical agitation during the most recent refueling outage did not correct the degraded condition associated with check valve 2FW-0191, but instead it masked it. The inspectors informed the licensee of their concerns, the licensee initiated Condition Report CR-2015-11013, and revised their operability determination documented in CR-2015-10961 to address these concerns.

Analysis.

The licensees failure to correct a condition adverse to quality was a performance deficiency. The performance deficiency was more than minor, and therefore a finding, because it was associated with the equipment performance attribute of the Mitigating Systems Cornerstone and affected the cornerstone objective to ensure availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the licensee failed to correct auxiliary feedwater check valve 2FW-0191 failure to seat in November 2012 resulting in an additional failure in November 2015. Using Inspection Manual Chapter (IMC) 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, dated June 19, 2012, inspectors determined that this finding was of very low safety significance (Green) because the finding

(1) was not a deficiency affecting the design and qualification of a mitigating structure, system, or component, and did not result in a loss of operability or functionality,

(2) did not represent a loss of system and/or function,

(3) did not represent an actual loss of function of at least a single train for longer than its allowed outage time, or two separate safety systems out-of-service for longer than their technical specification allowed outage time, and

(4) did not represent an actual loss of function of one or more non-technical specification trains of equipment designated as high safety-significant for greater than 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months in accordance with the licensees maintenance rule program. The finding has a problem identification and resolution cross-cutting aspect associated with evaluation, in that, the licensee failed to thoroughly evaluate issues to ensure that resolutions address extent of conditions. Specifically, the licensee failed to appropriately classify the issue of the check valve not seating and recognize this as a degraded condition [P.2].

Enforcement.

Title 10 CFR 50, Appendix B, Criterion XVI, Corrective Action, requires, in part, that measures shall be established to assure that conditions adverse to quality are promptly identified and corrected. Contrary to the above, prior to November 11, 2015, for quality related components associated with the auxiliary feedwater system, to which 10 CFR Part 50, Appendix B applies, the licensee failed to assure that conditions adverse to quality were promptly identified and corrected. Specifically, when a failure occurred during inservice testing of auxiliary feedwater check valves 2FW-0191 in November 2012, the licensee performed an operability evaluation of the auxiliary feedwater system and mechanically agitated the valve, but took no actions to identify and correct the degraded condition associated with the valve. In response to this issue, the licensee also performed an operability determination which established a reasonable expectation of operability pending implementation of corrective actions. This violation is being treated as an NCV, consistent with Section 2.3.2.a of the Enforcement Policy. The violation was entered into the licensees corrective action program as Condition Report CR-2015-11013. NCV 05000446/2016002-01, Failure to Correct Conditions Adverse to Quality

Cornerstone: Emergency Preparedness

1EP6 Drill Evaluation

Main article: IP 71114.06

.1 Emergency Preparedness Drill Observation

a. Inspection Scope

The inspectors observed an emergency preparedness drill on April 13, 2016, to verify the adequacy and capability of the licensees assessment of drill performance. The inspectors reviewed the drill scenario, observed the drill from the EOF, TSC, OSC, and simulator, and attended the post-drill critique. The inspectors verified that the licensees emergency classifications, off-site notifications, and protective action recommendations were appropriate and timely. The inspectors verified that any emergency preparedness weaknesses were appropriately identified by the licensee in the post-drill critique and entered into the corrective action program for resolution.

These activities constitute completion of one emergency preparedness drill observation sample, as defined in Inspection Procedure 71114.06.

b. Findings

No findings were identified.

RADIATION SAFETY

Cornerstones: Public Radiation Safety and Occupational Radiation Safety

2RS1 Radiological Hazard Assessment and Exposure Controls

Main article: IP 71124.01

a. Inspection Scope

The inspectors evaluated the licensees performance in assessing the radiological hazards in the workplace associated with licensed activities. The inspectors assessed the licensees implementation of appropriate radiation monitoring and exposure control measures for both individual and collective exposures. The inspectors walked down various portions of the plant and performed independent radiation dose rate measurements. The inspectors interviewed the radiation protection manager, radiation protection supervisors, and radiation workers. The inspectors reviewed licensee performance in the following areas:

Radiological hazard assessment, including a review of the plants radiological source term and associated radiological hazards. The inspectors also reviewed the licensees radiological survey program to determine whether radiological hazards were properly identified for routine and non-routine activities and assessed for changes in plant operations.

Instructions to workers, including labeling or marking containers of radioactive material, radiation work permits, actions for electronic dosimeter alarms, and changes to radiological conditions.

Contamination and radioactive material control including release of potentially contaminated material from the radiologically controlled area, radiological survey performance, radiation instrument sensitivities, material control and release criteria, and control and accountability of sealed radioactive sources.

Radiological hazards control and work coverage. During walk downs of the facility and job performance observations, the inspectors evaluated ambient radiological conditions, radiological postings, adequacy of radiological controls, radiation protection job coverage, and contamination controls. The inspectors also evaluated dosimetry selection and placement, and the use of dosimetry in areas with significant dose rate gradients. The inspectors examined the licensees controls for items stored in the spent fuel pool and evaluated airborne radioactivity controls and monitoring.

High radiation area and very high radiation area controls. During plant walk downs, the inspectors verified the adequacy of posting and physical controls, including for areas of the plan with the potential to become risk-significant high radiation areas.

Radiation worker performance and radiation protection technician proficiency with respect to radiation protection work requirements. The inspectors determined if workers were aware of the significant radiological conditions in their workplace, radiation work permit controls/limits in place, and were aware of their electronic alarming dosimeter dose and dose rate set points. The inspectors observed radiation protection technician job performance, including the performance of radiation surveys.

Problem identification and resolution for radiological hazard assessment and exposure controls. The inspectors reviewed audits, self-assessments, and corrective action program documents to verify problems were being identified and properly addressed for resolution.

These activities constitute completion of the seven required samples of radiological hazard assessment and exposure control program, as defined in Inspection Procedure 71124.01.

b. Findings

Introduction:

The inspectors reviewed a Green, self-revealed, non-cited violation of Technical Specification 5.7.1.e associated with the licensee allowing a worker access into a high radiation area without an adequate knowledge of the radiological conditions.

Specifically, the licensee briefed the worker on the conditions with outdated radiation survey information even though the 2-077-B penetration valve room was subject to changing radiological conditions. As a result, an individual entered areas with higher than expected dose rates.

Description:

On October 25, 2015, an individual received a high radiation area brief to enter the 2-077-B penetration valve room, which was posted as a high radiation area.

The radiation survey used to brief the individual to the expected radiological conditions was conducted on October 5, 2015, and indicated that the general area dose rates did not exceed approximately 50 millirem per hour. While performing residual heat removal (RHR) system venting in the valve room, the individual received a dose-rate alarm on his electronic dosimeter (ED). The individual left the area and notified radiation protection of the ED alarm. Follow-up radiation surveys by the licensee documented general area dose rates of approximately 210 millirem per hour.

Plant procedures allow the use of previously performed surveys to brief high radiation area entries, assuming there has been no movement or concentration of radioactive material or changes in plant conditions that could affect radiological conditions in the area. However, the 2-077-B penetration valve room contained RHR equipment that continuously cycled processed radioactive water. Additionally, the plant was in a refueling outage which affects radioactive material concentrations in the water running through the RHR system. The inspectors concluded that the radiological hazards present in the room were subject to change during RHR system operation and should have been evaluated consistent with that potential.

Analysis:

The inspectors determined that allowing worker access into a high radiation area (the 2-077-B penetration valve room) without an adequate knowledge of the radiological conditions was a performance deficiency. The inspectors determined that the performance deficiency was more than minor, and therefore a finding, because it affected the program and process attribute of the Occupational Radiation Safety cornerstone and adversely affected the cornerstone objective of ensuring adequate protection of worker health and safety from exposure to radiation. Specifically, because the licensee briefed the worker on the radiological conditions with outdated radiation survey information, the individual entered areas with higher than expected dose rates.

The finding was assessed using Inspection Manual Chapter 0609, Appendix C, Occupational Radiation Safety Significance Determination Process, issued August 19, 2008, and was determined to be of very low safety significance (Green) because the performance deficiency was not an ALARA planning issue, there was not an overexposure nor substantial potential for an overexposure, and the licensees ability to assess dose was not compromised.

The performance deficiency has a human performance cross-cutting aspect associated with work management, because the licensee failed to implement a process of planning, controlling, and executing work activities such that nuclear safety was the overriding priority. Specifically, the work process failed to identify and manage the radiological risk associated with an area of the plant that was known to be subject to changing radiological conditions based on equipment operations [H.5].

Enforcement:

Technical Specification 5.7.1.e requires, in part, that entry into high radiation areas be made only after dose rates in the area have been determined and entry personnel are knowledgeable of them. Contrary to the above, on October 25, 2015, an entry into a high radiation area was made before dose rates in the area had been determined and personnel were not knowledgeable of them. Specifically, the licensee briefed the worker on the conditions with outdated radiation survey information even though the 2-077-B penetration valve room was subject to changing radiological conditions. Since this violation was of very low safety significance and has been entered into the licensees corrective action program as Condition Report CR-2015-010211, this violation is being treated as a non-cited violation (NCV),consistent with Section 2.3.2.a of the Enforcement Policy. Corrective actions included performing follow-up radiation surveys and implementing improvements to the high radiation area access control program. NCV 0500445/2016002-03; 05000446/

2016002-02, Failure to Determine Dose Rates Prior to Allowing Entry into a High Radiation Area

2RS2 Occupational ALARA Planning and Controls

Main article: IP 71124.02

a. Inspection Scope

The inspectors assessed licensee performance with respect to maintaining individual and collective radiation exposures as low as is reasonably achievable (ALARA). The inspectors performed this portion of the attachment during the refueling outage, in order to directly observe the licensees ALARA process activities including planning, implementation of radiological work controls, execution of work activities, and ALARA review of work-in-progress. During the inspection the inspectors interviewed licensee personnel, reviewed licensee documents, and evaluated licensee performance in the following areas:

Radiological work planning, including work activities of exposure significance, and radiological work planning ALARA evaluations, initial and revised exposure estimates, and exposure mitigation requirements. The inspectors also verified that the licensees planning identified appropriate dose reduction techniques, reviewed any inconsistencies between intended and actual work activity doses, and determined if post-job (work activity) reviews were conducted to identify lessons learned. Specific work plans reviewed included the 1RF18 Reactor Coolant Pump Motor Swap and Pressurizer Work Activities.

Implementation of ALARA and radiological work controls including a review of the planned radiological administrative, operational, and engineering controls, compared to results achieved in the field.

Radiation worker performance including radiation protection technician performance during work activities performed in radiation areas, airborne radioactivity areas, or high radiation areas.

These activities constitute completion of two of the five required samples of occupational ALARA planning and controls program, as defined in Inspection Procedure 71124.02.

b. Findings

No findings were identified.

The inspectors verified that there were no unplanned exposures or losses of radiological control over locked high radiation areas and very high radiation areas during the period of October 1, 2015, to March 31, 2016. The inspectors reviewed a sample of radiologically controlled area exit transactions showing exposures greater than 100 millirem. The inspectors used definitions and guidance contained in Nuclear Energy Institute Document 99-02, Regulatory Assessment Performance Indicator Guideline, Revision 7, to determine the accuracy of the reported data.

These activities constitute verification of the occupational exposure control effectiveness performance indicator as defined in Inspection Procedure 71151.

b. Findings

.2 Annual Follow-up of Selected Issues

a. Inspection Scope

The inspectors selected one issue for an in-depth follow-up:

On November 9, 2015, during performance of protection channel time response testing on unit 1, a personnel error resulted in multiple alarms in the control, several instruments failing, and an unplanned reactivity change of approximately 1.5 percent power. The licensee identified that the maintenance personnel had used the incorrect test equipment because they did not recognize that a change to the procedure had been implemented that required a different test configuration.

The inspectors assessed the licensees problem identification threshold, cause analyses, extent of condition reviews and compensatory actions. The inspectors verified that the licensee appropriately prioritized the planned corrective actions and that these actions were adequate to correct the condition.

These activities constitute completion of one annual follow-up sample as defined in Inspection Procedure 71152.

b. Findings

No findings were identified.

.3 Semiannual Trend Review

a. Inspection Scope

The inspectors reviewed the licensees corrective action program and associated documents to identify trends that could indicate the existence of a more significant safety issue. The inspectors focused on the safety-related calculation change backlog. The inspectors reviewed documents and interviewed personnel to determine if the licensee completely and accurately identified problems in a timely manner commensurate with its significance, evaluated and dispositioned operability issues, considered the extent of condition, prioritized the problem commensurate with its safety significance, identified appropriate corrective actions, and completed corrective actions in a timely manner commensurate with the safety significance of the issue.

These activities constitute completion of one semi-annual trend review inspection sample as defined in Inspection Procedure 71152-05.

b. Findings

No findings were identified.

4OA5 Other Activities

a. Inspection Scope

The inspectors evaluated the impact of financial conditions on continued safe performance at Comanche Peak. In that the licensees parent company, Energy Future Holdings, was under bankruptcy protection/reorganization during the inspection period, NRC Region IV conducted special reviews of processes at Comanche Peak. The inspectors evaluated several aspects of the licensees operations to determine whether the financial condition of the station impacted plant safety. The factors reviewed included:

(1) impact on staffing,

(2) corrective maintenance backlog,

(3) changes to the planned maintenance schedule,

(4) corrective action program implementation, and

(5) reduction in outage scope, including risk-significant modifications. In particular, the inspectors verified that licensee personnel continued to identify problems at an appropriate threshold and enter these problems into the corrective action program for resolution. The inspectors also verified that the licensee continued to develop and implement corrective actions commensurate with the significance of the problems identified.

The special review of processes at Comanche Peak included continuous reviews by the Resident Inspectors, as well as the specialist-led baseline inspections completed during the inspection period which are documented previously in this report.

b. Findings

No findings were identified.

4OA6 Meetings, Including Exit

Exit Meeting Summary

On May 12, 2016, the inspectors presented the inspection results to Mr. J. Taylor, Director, Site Engineering, and other members of the licensee staff. The licensee acknowledged the issues presented. The licensee confirmed that any proprietary information reviewed by the inspectors had been returned or destroyed.

On May 13, 2016, the inspectors presented the radiation safety inspection results to Mr. J.

Dreyfuss, Plant Manager, and other members of the licensee staff. The licensee acknowledged the issues presented. The licensee confirmed that any proprietary information reviewed by the inspectors had been returned or destroyed.

On July 7, 2016, the resident inspectors presented the inspection results to Mr. T. McCool, Site Vice President, and other members of the licensee staff. The licensee acknowledged the issues presented. The licensee confirmed that any proprietary information reviewed by the inspectors had been returned or destroyed.

4OA7 Licensee-Identified Violations

The following violations of very low safety significance (Green) were identified by the licensee and are violations of NRC requirements which meet the criteria of the NRC Enforcement Policy for being dispositioned as non-cited violations.

Comanche Peak Unit 1, Operating License NPF-87, Condition 2.G, Fire Protection, requires, in part, that the licensee implement and maintain in effect all provisions of the approved fire protection program as described in the Final Safety Analysis Report through Amendment 78 and as approved in the Safety Evaluation Report and its supplements through Supplement 24.

Comanche Peak Unit 2, Operating License NPF-89, Condition 2.G, Fire Protection, requires, in part, that the licensee implement and maintain in effect all provisions of the approved fire protection program as described in the Final Safety Analysis Report through Amendment 87 and as approved in the Safety Evaluation Report and its supplements through Supplement 27.

The stations approved fire protection program includes Fire Protection Report, Revision 29, Section 5.3.8, Fire Area EO - Control Room, includes Deviation 3c-1, Control Room Missile Door, which requires, in part, that since the control room missile door in the west wall is not a three hour rated fire door, the area of the turbine deck within 100 feet of the door is to be void of combustibles. Contrary to the above, on May 5, 2016, the licensee stored combustible materials within 100 feet of the control room missile door in the west wall. Specifically, licensee personnel identified that contractors had stored combustibles within the combustible free zone, and that no compensatory measures had been implemented for the deviation from the Fire Protection Report. The licensee implemented a periodic roving fire watch to compensate for the reduction in fire protection. The violation is more than minor because if left uncorrected, it could lead to a more significant safety concern. Using Inspection Manual Chapter 0609, 04, Initial Characterization of Findings, and Inspection Manual Chapter 0609, Appendix F, Fire Protection Significance Determination Process, the inspector determined that the violation is of very low safety significance (Green) because the finding did not affect the ability of either unit to achieve safe shutdown. The violation was entered into the licensees corrective action program as CR-2016-004167.

Technical Specification 5.4.1.a states, in part, that written procedures shall be established, implemented, and maintained covering the applicable procedures recommended in Regulatory Guide 1.33, Revision 2, Appendix A, February 1978.

Regulatory Guide 1.33, Appendix A, Section 9.a., identifies procedures for maintenance as required procedures. Work order 4831032 is a procedure established by the licensee for performing maintenance on diesel generator 2-02. The work order provided instructions for installation of the magnetic speed pickup sensor cable. Contrary to the above, from October 1996 through March 2, 2016, the licensee failed to install the unit 2 diesel generator 2-02 magnetic speed pickup sensor cable in accordance with the approved instructions. Specifically, the speed sensor cable conduit was not fully threaded onto the cable plug. This inadequate installation was present until 2016, when the conduit threaded connection was physically impacted at an undetermined time. The impact caused the conduit connection to break and the conduit to separate from the plug, leaving the cable leads exposed but intact. A licensee technician identified the broken connection during a system walk down on March 2, 2016. The licensee declared the diesel generator inoperable and restored the cable to its design configuration. The licensee analyzed the apparent thread engagement, and determined that, prior to the break in the conduit connection, the cable would have maintained its function in a seismic event, but after the break, the cable function could not be assured. The licensee determined that a failure of the cable would result in the diesel generator exceeding its allowed frequency, but would not result in a diesel generator failure to run. Because the time that the break occurred could not be determined, the diesel generator was assumed to be inoperable at the time of discovery. The violation is more than minor because it affected the configuration control attribute of the Mitigating Systems cornerstone and impacted the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using Inspection Manual Chapter 0609, Attachment 04, Initial Characterization of Findings, and Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process for Findings At-Power, the inspector determined that the violation is of very low safety significance (Green) because the finding did not represent a loss of system or function, and did not represent a loss of function of a single train for greater than its technical specification allowed outage time. The violation was entered into the licensees corrective action program as CR-2016-001941.