IR 05000445/2017004

NRC Integrated Inspection Report 05000445/2017004 and 05000446/2017004

ML18040A634
Person / Time
Site:	Comanche Peak
Issue date:	02/09/2018
From:	Mark Haire NRC Region 4
To:	Peters K Vistra Operations Company
Mark Haire
References
IR 2017004
Download: ML18040A634 (47)

Text

February 9, 2018 Ken J. Peters, Senior Vice President and Chief Nuclear Officer Vistra Operations Company LLC P.O. Box 1002 Glen Rose, TX 76043 SUBJEC T: COMANCHE PEAK NUCLEAR POWER PLANT - NRC INTEGRATED INSPECTION REPORT 05000445/2017004 and 05000446/2017004

Dear Mr. Peters:

O n December 31, 2017, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at your Comanche Peak Nuclear Power Plant, Units 1 and 2. On January 3, 201 8 , the NRC inspectors discussed the results of this inspection with you and other members of your staff. The results of this inspection are documented in the enclosed report.

NRC inspectors documented six findings of very low safety significance (Green) in this report. All of these findings involved violations 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. 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; and the NRC resident inspector at the Comanche Peak Nuclear Power Plant. 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 U.S.

Nuclear Regulatory Commission, ATTN: Document Control Desk, Washington, DC 20555-0001; with copies to the Regional Administrator, Region IV; and the NRC resident inspector at the Comanche Peak Nuclear Power Plant.

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 made available electronically for public inspection in the NRC's Public Document Room or the NRC's Agencywide Documents Access and Management System (ADAMS), accessible from the NRC Web site at http://www.nrc.gov/reading-rm/adams.html. UNITED STATES NUCLEAR REGULATORY COMMISSION REGION IV 1600 E. LAMAR BLVD ARLINGTON, TX 76011-4511 K.Peters2 To the extent possible, your response, if any, should not include any personal privacy, proprietary, or safeguards information so that it can be made available to the public without redaction.

Sincerely,

/RA/ Mark S. Haire, Chief Project Branch A Division of Reactor Projects Docket Nos. 5000445 and 5000446 License Nos. NPF-87 and NPF-89 Enclosure: Inspection Report 05000445/201 7004 and 05000446/2017004 w/ Attachments: 1.) Supplemental Information 2.) Document Request

K.Peters3 COMANCHE PEAK NUCLEAR POWER PLANT - NRC INTEGRATED INSPECTION REPORT 05000445/2017004 and 05000446/2017004 - Dated February 9, 2018 DISTRIBUTION KKennedy, RA SMorris, DRA TPruett, DRP AVegel, DRS JClark, DRS RLantz, DRP JJosey, DRP RKumana, DRP MHaire, DRP RAlexander, DRP DLackey, DRP JBowen, RIV/OEDO SKirkwood, RC VDricks, ORA JWeil, OCA MO'Banion, NRR AMoreno, RIV/OCA BMaier, RSLO THipschman, IPAT EUribe, IPAT MHerrera, DRMA RIV ACES ROP Reports Electronic Distribution for Comanche Peak Nuclear Power Plant

ADAMS ACCESSION NUMBER: SUNSI ReviewBy: MSH/dllADAMS Yes NoNon-SensitiveSensitivePublicly AvailableNon-Publicly AvailableOFFICE SRI:DRP/A RI:DRP/A SPE:DRP/A BC:EB1 BC:EB2 BC:OB NAME JJosey RKumana RAlexander TFarnholtz GWerner VGaddy SIGNATURE /RA/ /RA/ /RA/ /R A/ /RA/JMM for

/RA/ DATE 2/8/18 2/8/18 02/07/18 02/07/2018 02/07/2018 2/7/18 OFFICE BC:PSB2 TL-IPAT BC:DRP/A NAME HGepford THipschman MHaire SIGNATURE /RA/ /RA/HAF for

/RA/ DATE 02/07/18 02/08/2018 2/9/18 Enclosure U.S. NUCLEAR REGULATORY COMMISSI ON REGION IV Docket: 05000445 , 05000446 License: NPF-87 , NPF-89 Report: 05000445/2017004 and 05000446/2017004 Licensee: Vistra Operations Company, LLC Facility: Comanche Peak Nuclear Power Plant, Units 1 and 2 Location: 6322 N. FM-56, Glen Rose, Texas Dates: October 1 through December 31, 2017 Inspectors:

J. Josey, Senior Resident Inspector R. Kumana, Resident Inspector J. Drake, Senior Reactor Inspector L. Carson, II, Senior Health Physicist S. Money, Health Physicist Approved By: Mark S. Haire Chief Project Branch A Division of Reactor Projects

2

SUMMARY

IR 05000445/2017004; 05000446/2017004; 10/01/2017 - 12/31/2017; Comanche Peak Nuclear Power Plant; Equipment Alignment; Inservice Inspection Activities; Maintenance Effectiveness; Maintenance Risk Assessments and Emergent Work Control; Operability Determinations and Functionality Assessments; Refueling and Other Outage Activities

The inspection activities described in this report were performed between October 1 and December 31, 2017, by the resident inspectors at Comanche Peak Nuclear Power Plant and inspectors from the NRC's Region IV office. Six findings of very low safety significance (Green) are documented in this report. All of these findings involved violations of NRC requirements. The significance of inspection findings is indicated by their color (i.e., Green, greater than Green, White, Yellow, or Red), determined using Inspection Manual Chapter 0609, "Significance Determination Process," dated April 29, 2015. Their cross-cutting aspects are determined using Inspection Manual Chapter 0310, "Aspects within the Cross-Cutting Areas," dated December 4, 2014. Violations of NRC requirements are dispositioned in accordance with the NRC Enforcement Policy. The NRC's program for overseeing the safe operation of commercial nuclear power reactors is described in NUREG

-1649, "Reactor Oversight Process," dated July 2016.

Cornerstone: Initiating Events

Green.

The inspectors identified a non-cited violation of 10CFR50.55a(g)(4) involving the licensee's failure to perform pressure testing of all Class 1 pressure retaining components for Units 1 and 2, in accordance with the applicable edition of Section XI of the ASME Boiler and Pressure Vessel Code. Specifically, prior to November 1, 2017, the licensee failed to perform the required pressure test on th ose portions of the Class 1 pressure boundary between the first and second isolation valves in the injection and return path of safety-related systems for both units in accordance with ASME Code,Section XI, Article IWB

-5000, "System Pressure Tests

." This issue does not represent an immediate safety concern because the licensee performed an operability evaluation for Unit 2 which demonstrated a reasonable expectation of operability, and received an approved relief request prior to Unit 1 startup. This finding was entered into the licensee's corrective action program as Condition Reports CR

-2017-010530 and C R-2017-011968. The inspectors determined that the licensee's failure to perform a system leakage test of all Class 1 pressure retaining components for each of the units was a performance deficiency. This finding was more than minor because it affected the initiating events cornerstone attribute of equipment reliability

, and affected the cornerstone objective to limit the likelihood of events that upset plant stability and challenge critical safety functions. In accordance with Inspection Manual Chapter 0609, Appendix A, "The Significance Determination Process for Findings At-Power," dated June 19, 2012, Exhibit 1 , "Initiating Events Screening Questions ," the finding was determined to be of very low safety significance (Green) because the finding did not result in exceeding the reactor coolant system leak-rate for a small loss-of-coolant accident and did not affect other systems used to mitigate a loss-of-coolant accident resulting in a total loss of their function. The finding has a problem identification and resolution cross-cutting aspect associated with evaluation. Although the licensee identified that they had failed to properly perform pressure testing on portions of various ASME Code Class 1 systems, they required prompting from the inspectors to thoroughly evaluate the issue to ensure that their resolution fully addressed ASME requirements and the need to obtain a relief request prior to restarting Unit 1 following the outage [P.2]. (Section 1R08.5)3

Cornerstone: Mitigating Systems

Green.

The inspectors identified a non-cited violation of 10 CFR 50, Appendix B, Criterion XVI, "Corrective Action," associated with the licensee's failure to take timely corrective action s for a condition adverse to quality.

Specifically, since 2010 the licensee failed to take corrective actions for indications of degraded concrete in the service water intake structure. This issue does not represent an immediate safety concern because the licensee performed detailed engineering evaluations and tests to confirm the structural integrity of the intake structure, and will implement corrective actions to repair the affected concrete in the future. The licensee entered this issue into the corrective action program as Condition Report CR

-2018-000088. The licensee's failure to take timely and adequate corrective actions to correct a condition adverse to quality was a performance deficiency. The performance deficiency is more than minor , and therefore a finding, because it is 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 failure to correct the degraded condition resulted in continued degradation with no evaluation to determine continu ed operability. Using Inspection Manual Chapter 0609, Attachment 04, "Initial Characterization of Findings," dated October 7, 2016, and Inspection Manual Chapter 0609, Appendix A , "Significance Determination Process for Findings At-Power," Exhibit 4 , "External Events Screening Questions," the inspectors determined the finding was of very low safety significance (Green) because the finding is a deficiency affecting the design or qualification of a mitigating SSC, but the SSC maintain ed its operability. The finding has a human performance cross-cutting aspect associated with conservative bias, in that, the licensee failed to ensure that the individuals used decision making practices that emphasized prudent choices [H.14

]. (Section 1R12)

Green.

The inspectors identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, associated with the licensee's failure to accomplish activities affecting quality in accordance with documented procedures. Specifically, the licensee breached control room door E

-40A, a tornado missile barrier, without following the requirements of standing order OSO

-008. This issue does not represent an immediate safety concern because

, when this was identified

, the licensee took action to immediately close door E

-40A. The licensee entered this issue into the corrective action program as Condition Report CR

-2017-010661. The licensee's failure to follow the requirements of Standing Order OSO

-008 when breaching door E

-40A was a performance deficiency. This performance deficiency is more than minor, and therefore a finding, because it is associated with protection against external factors 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.

Using Inspection Manual Chapter 0609, Attachment 04, "Initial Characterization of Findings," dated October 7, 2016, and Inspection Manual Chapter 0609, Appendix A, "Significance Determination Process for Findings At-Power," Exhibit 2 , "Mitigating Systems Screening Questions," the inspectors determined the finding was of very low safety significance (Green) because: (1) it was not a design deficiency; (2) it did not represent a loss of system and/or function; (3) it did not represent an actual loss of function of at least a single train for longer than its technical specification allowed outage time; (4) and it did not result in the loss of a high safety significant non-technical specification train. The finding has a problem identification and resolution cross-cutting aspect associated with resolution, in that the licensee failed to take adequate corrective actions from a previous issue with this door [P.3]. (Section 1R13)

Cornerstone: Barrier Integrity

Green.

The inspectors identified a non-cited violation of 10 CFR 50, Appendix B, Criterion V , "Instructions, Procedures, and Drawings," associated with the licensee's failure to prescribe adequate procedures for closing the containment equipment hatch for both units. Specifically, the licensee's procedure failed to specify the correct bolting arrangement and torque for closure during emergency conditions. This issue does not represent an immediate safety concern because at the time of discovery rapid containment closure was not required, and the licensee implemented corrective actions to change the procedure to reflect the correct torque value and bolting pattern

. The licensee entered this issue into the corrective action program as Condition Report CR

-2017-011236. The licensee's failure to prescribe adequate procedures to perform quality related activities was a performance deficiency. The performance deficiency is more than minor , and therefore a finding, because it is associated with the procedure quality attribute of the Barrier Integrity Cornerstone and affected the cornerstone objective to provide reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. Specifically, the inadequate procedure resulted in a potentially degraded containment barrier in the event of core damage during shutdown conditions. Using Inspection Manual Chapter 0609, Attachment 04, "Initial Characterization of Findings," dated October 7, 2016, and Inspection Manual Chapter 0609, Appendix G, Attachment 1, "Shutdown Operations Significance Determination Process Phase 1 Initial Screening and Characterization of Findings," Exhibit 4, "Barrier Integrity Screening Questions," the inspectors determined the finding degraded the ability to close or isolate containment and required evaluation under Inspection Manual Chapter 0609, Appendix H, "Containment Integrity Significance Determination Process," dated May 6, 2004. Using Large Early Release Frequency (LERF) type screening process, the inspectors determined the finding was a "Type B LERF" finding because the finding did not affect core damage frequency, and that a Phase 2 estimate was required because the containment equipment hatch was important to LERF in accordance with Table 6.1. The inspectors used Table 6.2 Phase 2 Risk Significance to determine the finding was of very low safety significance (Green) because it did not meet the threshold for a Whi te finding for leakage from containment to the environment being greater than 100 percent containment volume per day through containment penetration seals, isolation valves or vent and purge systems.

The finding was not assigned a cross-cutting aspect because the performance deficiency was not reflective of current performance

. (Section 1R04)

Green.

Inspectors identified a non-cited violation of 10 CFR 50, Appendix B, Criterion V, "Instructions, Procedures, and Drawings," associated with the licensee's failure to follow the requirements of station procedure STI

-606.01, Work Control Process, Revision 5. Specifically, the work order for emergent maintenance activities on Containment Access Corridor door S1

-35G was not reviewed by operations for plant impacts prior to work commencing. As a result, S1-35G was blocked open with the Unit 1 containment personnel airlock and equipment hatch open, causing the Unit 2 primary plant ventilation system to be inoperable.

This issue does not represent an immediate safety concern because the licensee took action to enter the required technical specification action statement and establish appropriate compensatory measures for the activity.

The licensee entered this issue into the corrective action program as Condition Report CR-2017-011424. The licensee's failure to follow procedural requirements and review work activities for plant impact when blocking open door S1

-35G for emergent maintenance was a performance deficiency. This performance deficiency is more than minor, and therefore a finding, because it is associated with the SSC and barrier performance attribute of the Barrier Integrity cornerstone

, and adversely affected the cornerstone objective to provide reasonable assurance that physical design barriers (fuel cladding, reactor coolant system, and containment) protect the public from radionuclide releases caused by accidents or events. Specifically, with the Unit 1 containment personnel airlock and equipment hatch open and door S1

-35G blocked open

, the Unit 2 primary plant ventilation system could not perform its function. Using Inspection Manual Chapter 0609, Attachment 04, "Initial Characterization of Findings," dated October 7, 2016, and Inspection Manual Chapter 0609, Appendix A, "Significance Determination Process for Findings At-Power," Exhibit 4, "External Events Screening Questions," the inspectors determined the finding to be of very low safety significance (Green) because it only represented a degradation of the radiological barrier function provided for the auxiliary building secondary containment. The finding has a human performance cross-cutting aspect associated with team work, in that the licensee failed to ensure that work groups adequately communicate and coordinate their activities across organizational boundaries to ensure that nuclear safety is maintained [H.4].

(Section 1R15)

Green.

The inspectors identified a non-cited violation of 10 CFR 26.205 associated with the licensee's failure to calculate and control work hours of personnel subject to work hour controls. Specifically, the licensee failed to track work hours for personnel assigned to emergency containment equipment hatch closure duty, resulting in multiple violations of work hour requirements. This issue does not represent an immediate safety concern because at the time of discovery no personnel were in excess of work hour requirements, and the licensee took action to include the individuals in work hour controls during future outages. The licensee entered this issue into the corrective action program as Condition Report CR-2017-012922. The licensee's failure to control work hours for personnel subject to work hour controls was a performance deficiency. The performance deficiency is more than minor , and therefore a finding, because it is associated with the human performance attribute of the Barrier Integrity Cornerstone and affected the cornerstone objective to provide reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accident s or events. Specifically, the failure to control work hours for personnel assigned to emergency containment hatch closure introduced a higher risk of human performance errors in maintaining containment integrity. Using Inspection Manual Chapter 0609, Attachment 04, "Initial Characterization of Findings," dated October 7, 2016, and Inspection Manual Chapter 0609, Appendix G, Attachment 1, "Shutdown Operations Significance Determination Process Phase 1 Initial Screening and Characterization of Findings," Exhibit 4, "Barrier Integrity Screening Questions," the inspectors determined the finding was of very low safety significance because it did not degrade the ability to close or isolate containment and did not degrade the physical integrity of reactor containment. The finding was not assigned a cross-cutting aspect because the cause of the performance deficiency occurred more than three years prior and therefore is not reflective of current performance.

(Section 1R20)

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PLANT STATUS

Unit 1 began the inspection period at approximately 100 percent power

. On October 8, 2017, Unit 1 was shut down for a planned refueling outage. Unit 1 returned to full power on November 11, 2017. Unit 1 operated at full power for the rest of the inspection period.

Unit 2 began the inspection period at approximately 100 percent power. On November 25, 2017, Unit 2 mas manually tripped due to a simultaneous loss of both main feed water pump s. Unit 2 returned to full power on December 3, 2017. Unit 2 operated at full power for the rest of the inspection period.

REPORT DETAILS

REACTOR SAFETY

Cornerstones: Initiating Events, Mitigating Systems, and Barrier Integrity 1 R 01 Adverse Weather Protection (71111.01)

.1 Readiness for Seasonal Extreme Weather Conditions

a. Inspection Scope

On December 26

, 2017, the inspectors completed an inspection of the station's readiness for seasonal extreme weather conditions. The inspectors reviewed the licensee's adverse weather procedures for cold weather and evaluated the licensee's implementation of these procedures. The inspectors verified that prior to the onset of cold weather, the licensee had corrected weather

-related equipment deficiencies identified during the previous cold weather season

. The inspectors selected two risk-significant systems that were required to be protected from cold weather

emergency diesel generators fire protection The inspectors reviewed the licensee's procedures and design information to ensure the systems or components would remain functional when challenged by adverse weather. The inspectors verified that operator actions described in the licensee's procedures were adequate to maintain readiness of these systems. The inspectors walked down portions of these systems to verify the physical condition of the adverse weather protection features. These activities constituted one sample of readiness for seasonal adverse weather, as defined in Inspection Procedure 71111.01.

b. Findings

No findings were identified.

.2 Readiness to Cope with External Flooding

a. Inspection Scope

On December 20, 2017, the inspectors completed an inspection of the station's readiness to cope with external flooding. After reviewing the licensee's flooding analysis, the inspectors chose two plant areas that were susceptible to flooding:

safety chiller rooms uninterruptable power supply heating , ventilation, and air conditioning rooms The inspectors reviewed plant design features and licensee procedures for coping with flooding. The inspectors walked down the selected areas to inspect the design features, including the material condition of seals, drains, and flood barriers. The inspectors evaluated whether credited operator actions could be successfully accomplished.

These activities constituted one sample of readiness to cope with external flooding, as defined in Inspection Procedure 71111.01.

b. Findings

No findings were identified.

1 R 04 Equipment Alignment (71111.04)

.1 Partial Walk

-Down

a. Inspection Scope

The inspectors performed partial system walk

-downs of the following risk

-significant systems: October 15, 2017 , Unit 1, containment hatches during refueling operations November 7, 2017, Unit 1 , train A containment spray system following startup The inspectors reviewed the licensee's 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 sample s as defined in Inspection Procedure 71111.04.

b. Findings

Introduction.

The inspectors identified a Green , non-cited violation of 10 CFR 50, Appendix B, Criterion V , "Instructions, Procedures, and Drawings," associated with the licensee's failure to prescribe adequate procedures for closing the containment equipment hatch for both units

. Specifically, the licensee's procedure failed to specify the correct bolting arrangement and torque for closure during emergency conditions.

Description.

During the Unit 1 refueling outage 1RF19, the inspectors observed the licensee's training on rapid containment equipment hatch closure in response to abnormal events. The inspectors also reviewed the licensee's procedures and supporting calculations. The inspectors noted that the engineering analysis used to demonstrate successful closure of containment in response to a refueling accident or loss of shutdown cooling made several assumptions to demonstrate adequate clamping force for the equipment hatch. The analysis assumed that 4 of the 16 bolts would be tightened, and that the 4 bolts selected would be 90 degrees apart. The analysis also assumed that the bolts would be tightened to a minimum preload value that the licensee defined as "snug tight

." The inspectors noted that the procedure STI 600.01, "Inspecting Plant Equipment and Sensitive Equipment Controls," Revision 1, did not require the bolts to be tightened to "snug tight" and also did not require the four bolts to be spaced 90 degrees apart. Based on their observations of containment closure drills, the inspectors determined that the licensee was not selecting bolts 90 degrees apart. Instead, the bolts were approximately 67.5 degrees apart on the vertical axis and 112.5 degrees apart on the horizontal axis. The inspectors also determined that the personnel tightening the bolts were not verifying sufficient load on the bolts was being applied. The inspectors concluded that the licensee had not established adequate procedures for emergency closure of the containment equipment hatch.

The licensee entered this into the corrective action program as Condition Report CR-2017-011236. At the time the inspectors raised the concern, the time to boil on Unit 1 was sufficiently long that the licensee would have had time to properly close the hatch. The inspectors noted that earlier in the outage, the licensee would have improperly closed the hatch in the event of a loss of shutdown cooling. The licensee determined that the closure method used would be acceptable to ensure protecti on against a fuel handling accident.

The licensee performed an evaluation to determine whether the containment hatch would have prevented a release in the event of a loss of shutdown cooling. The licensee determined that there was sufficient margin in the bolt stress to ensure containment with the specific asymmetric spacing that was utilized during the outage.

However, the licensee was unable to demonstrate that the procedure would result in sufficient preload on the four bolts used for closure. The inspectors determined that the issue could have resulted in an inadequate seal in the event containment closure was required, but that the resulting leakage would be restricted to the gasket area. The licensee determined, assuming no preload on the bolts, that the worst case leakage would be bounded by 30 percent containment volume turnover per day.

Analysis.

The licensee's failure to prescribe adequate procedures to perform quality related activities was a performance deficiency. The performance deficiency is more than minor , and therefore a finding, because it is associated with the procedure quality attribute of the Barrier Integrity Cornerstone and affected the cornerstone objective to provide reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. Specifically, the inadequate procedure resulted in a potentially degraded containment barrier in the event of core damage during shutdown conditions. Using Inspection Manual Chapter 0609, Attachment 04, "Initial Characterization of Findings," dated October 7, 2016, and Inspection Manual Chapter 0609, Appendix G, Attachment 1, "Shutdown Operations Significance Determination Process Phase 1 Initial Screening and Characterization of Findings," Exhibit 4, "Barrier Integrity Screening Questions," the inspectors determined the finding degraded the ability to close or isolate containment and required evaluation under Inspection Manual Chapter 0609, Appendix H, "Containment Integrity Significance Determination Process," dated May 6, 2004. Using the Large Early Release Frequency (LERF) type screening process, the inspectors determined that the issue was a "Type B LERF" finding because it did not affect core damage frequency, and that a Phase 2 estimate was required because the containment equipment hatch was important to LERF in accordance with Table 6.1. The inspectors used Table 6.2 Phase 2 Risk Significance to determine the finding was of very low safety significance (Green) because it did not meet the threshold for a White finding for leakage from containment to the environment because the estimated leakage was determined to be less than 100 percent containment volume turnover per day through containment penetration seals, isolation valves or vent and purge systems.

The inspectors determined that the procedure used for containment hatch closure had not been updated in more than 3 years therefore the finding was not assigned a cross

-cutting aspect because the performance deficiency was not reflective of current performance

.

Enforcement.

10 CFR 50, Appendix B, Criterion V, "Instructions, Procedures, and Drawings," requires, in part, that activities affecting quality shall be prescribed by procedures of a type appropriate to the circumstances. Contrary to the above, from July 2015 through October 2017, the licensee failed to prescribe activities affecting quality by procedures of a type appropriate to the circumstances.

This issue does not represent an immediate safety concern because

, at the time of discovery

, rapid containment closure was not required, and the licensee generated corrective actions to modify their procedures. Since this violation was of very low safety significance (Green) and has been entered into the corrective action program as Condition Report CR-2017-011236, this violation is being treated as a non

-cited violation consistent with Section 2.3.2.a of the NRC Enforcement Policy.

(NCV 05000445/2017004

-01; 05000446/2017004

-01, Inadequate Procedure for Containment Closure)1 R 05 Fire Protection (71111.05)

.1 Quarterly Inspection

a. Inspection Scope

The inspectors evaluated the licensee's fire protection program for operational status and material condition. The inspectors focused their inspection on four plant areas important to safety: November 20, 2017, fire area EL62, Unit 1 stairwell November 20, 2017, fire area AF33, Unit 1 component cooling water pump 1

-01 November 2 2, 2017, fire area 2SH11, Unit 2 diesel generator day tank November 22, 2017, fire area SB2, Unit 1 containment spray pumps For each area, the inspectors evaluated the fire plan against defined hazards and defense-in-depth features in the licensee's 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.

1 R 06 Flood Protection Measures (71111.06)

a. Inspection Scope

On December 19 , 2017, the inspectors completed an inspection of the station's ability to mitigate flooding due to internal causes. After reviewing the licensee's flooding analysis, the inspectors chose two plant areas containing risk

-significant structures, systems, and components that were susceptible to flooding:

Unit 1, train A containment spray pump room Unit 1, train B containment spray pump room The inspectors reviewed plant design features and licensee procedures for coping with internal flooding. The inspectors walked down the selected areas to inspect the design features, including the material condition of seals, drains, and flood barriers. The inspectors evaluated whether operator actions credited for flood mitigation could be successfully accomplished.

These activities constituted completion of one flood protection measures sample as defined in Inspection Procedure 71111.06.

b. Findings

No findings were identified.

1 R 07 Heat Sink Performance (71111.07)

a. Inspection Scope

On November 17, 2017, 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 Electric Power Research Institute (EPRI) NP-7552 for the Unit 1 containment spray pump 1

-01 lube oil cooler heat exchanger. Additionally, the inspectors walked down the Unit 1 containment spray pump 1-01 lube oil cooler heat exchanger to observe its performance and material condition and verified that the heat exchanger was correctly categorized under the Maintenance Rule and was receiving the required maintenance.

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

b. Findings

No findings were identified.

1 R 08 Inservice Inspection Activities (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 COMPONENT IDENTIFICATION EXAMINATION TYPE Safety Injection TBX-1-4303-H 3 Visual Test

- 3 Safety Injection TBX-1-4201-H 3 Visual Test

- 3 Safety Injection TBX-1-4303-H 3 Penetrant Test Safety Injection TBX-1-4201-H 4 Penetrant Test Steam Generator TBX-1-3100-3-1 Ultrasonic Test Pressurizer TBX-1-2100-6 Ultrasonic Test Pressurizer TBX-1-2100-10 Magnetic Particle Test The inspectors reviewed records for the following nondestructive examinations:

SYSTEM COMPONENT IDENTIFICATION EXAMINATION TYPE Reactor Coolant TBX-1-4500-H 1 Visual Test

- 3 Safety Injection TBX-1-4301-H 1 Visual Test

- 3 Safety Injection TBX-1-4301-H 2 Visual Test

- 3 Residual Heat Removal TBX-2-2530-H 11-H 1 Visual Test

- 3 Component Cooling CC-1-SB-006-H 1 Visual Test

- 3 Safety Injection TBX-1-4306-H 28 Visual Test

- 3 Building and Structures Equipment Hatch Visual Test

- 3 Reactor Coolant TBX-1-1300A-Sup Visual Test

- 3 Reactor Coolant TBX-1-4500-H1 Ultrasonic Test Pressurize r TBX-1-4504-1 Ultrasonic Test During the review and observation of each examination, the inspectors observed that activities were performed in accordance with the American Society of Mechanical Engineers (ASME) Code requirements and applicable procedures. The inspectors also reviewed the qualifications of nondestructive examination technicians performing inspections and determined they were current.

The inspectors directly observed a portion of the following welding activities:

SYSTEM WELD IDENTIFICATION WELD PROCESS Reactor Coolant 1 CS-0016 FW-1-3 A Gas Tungsten Arc Welding Reactor Coolant 1 CS-0017 FW-1-3 A Gas Tungsten Arc Welding The inspectors reviewed records for the following welding activities:

SYSTEM WELD IDENTIFICATION WELD PROCESS Service Water Weld 17-3 Gas Tungsten Arc Welding Auxiliary Feedwater Weld FW-44 Gas Tungsten Arc Welding Auxiliary Feedwater Weld FW-45 Gas Tungsten Arc Welding Auxiliary Feedwater Weld A Gas Tungsten Arc Welding Station Service Water CP1-SWSRPL-03 Weld Repair of Sealing Surface Gas Tungsten Arc Welding The inspectors reviewed that the welding procedure specifications and the welders were properly qualified in accordance with ASME Code Section IX requirements. The inspectors also determined that essential variables were identified, recorded in the procedure qualification record, and formed the bases for qualification of the welding procedure specifications.

b. Findings

No findings were identified.

.2 Vessel Upper Head Penetration Inspection Activities

The bare metal visual inspection of the reactor vessel upper head penetrations was not performed in this outage.

.3 Boric Acid Corrosion Control Inspection Activities

a. Inspection Scope

The inspectors reviewed the licensee's implementation of its boric acid corrosion control program for monitoring degradation of those systems that could be adversely affected by boric acid corrosion. The inspectors reviewed the documentation associated with the licensee's boric acid corrosion control walkdown as specified in Procedure STA

-737, "Boric Acid Corrosion Detection and Evaluation," Revision 8. The inspectors verified that the visual inspections emphasized locations where boric acid leaks could cause degradation of safety

-significant components, and that engineering evaluations used corrosion rates applicable to the affected components and properly assessed the effects of corrosion induced wastage on structural or pressure boundary integrity. The inspectors observed that corrective actions taken were consistent with the ASME Code and 10 CFR 50, Appendix B requirements.

b. Findings

No findings were identified.

.4 Steam Generator Tube Inspection Activities

a. Inspection Scope

The inspectors reviewed the steam generator tube eddy current examination scope and expansion criteria, and determined that these criteria met technical specification requirements, EPRI guidelines, and commitments made to the NRC. The inspectors also verified that the eddy current examination inspection scope included areas of degradations that were known to represent potential eddy current test challenges such as the top of tube sheet, tube support plates, and U

-bends. The inspectors confirmed that no repairs were required at the time of the inspection.

Steam Generator Inspection The inspectors verified that the number and sizes of steam generator tube flaws/degradation identified were consistent with the licensee's previous outage operational assessment predictions.

The inspectors verified that steam generator eddy current examination scope and expansion criteria met technical specification requirements.

The inspectors verified that eddy current probes and equipment configurations used to acquire data from the steam generator tubes were qualified to detect the known/expected types of steam generator tube degradation in accordance with Appendix H, "Performance Demonstration for Eddy Current Examination of EPRI Document 1013706."

The inspectors reviewed the licensee's identification of the following tube degradation mechanisms:

t ube support plate wear foreign object wear Secondary Side Inspections The inspectors reviewed the completed portions of secondary side inspection results and verified the licensee took or planned to take corrective actions in response to the observed degradation.

The inspectors reviewed the licensee's planned actions in response to several foreign objects that had been identified on the secondary side of the steam generators.

b. Findings

No findings were identified.

.5 Identification and Resolution of Problems

a. Inspection Scope

The inspectors reviewed 25 condition reports which dealt with inservice inspection activities and found the corrective actions for inservice inspection issues were appropriate. From this review the inspectors concluded that the licensee has an appropriate threshold for entering inservice inspection issues into the corrective action program and has procedures that direct a root cause evaluation when necessary. The licensee also has an effective program for applying industry inservice inspection operating experience. Specific documents reviewed during this inspection are listed in the attachment.

b. Findings

Introduction.

The inspectors identified a Green

, non-cited violation of 10CFR50.55a(g)(4) involving the licensee's failure to perform a system pressure test of all Class 1 pressure retaining components for Units 1 and 2, in accordance with the applicable edition of Section XI of the ASME Boiler and Pressure Vessel Code. Specifically, prior to November 1, 2017, the licensee failed to perform the required pressure test of the portions of the Class 1 pressure boundary between the first and second isolation valves in the injection and return path of safety

-related systems for both units in accordance with ASME Code Section XI, Article IWB

-5000 requirements.

Description.

During an inservice inspection program review, the inspectors noted that the licensee had identified that they were not correctly performing system leakage tests of all Class 1 pressure retaining components for each of the units. The inspectors identified, through further review and discussion, that the licensee had not adequately addressed the implications of this for either unit. Unit 1 was in a refueling outage, and Unit 2 was operating at 100 percent reactor power. Specifically, the licensee was performing the system leakage tests for the portions of the Class 1 boundary between the first and second isolation valves in the injection and return path of interfacing safety

-related systems either during shutdown activities or prior to reaching a pressure corresponding to 100 percent reactor power (2235 psig +/

- 15 psig). Article IWB

-5000, "System Pressure Tests," of Section XI of the ASME Code requires that all pressure retaining components be pressure tested via a system leakage test per IWB

-5220, "System Leakage Test." The licensee is required by 10 CFR 50.55a(g)(4)to comply with the requirements imposed by Section XI of the ASME Code or request exemption from particular requirements via a relief request. The inspectors had additional concerns with the licensee's evaluation that a relief request was not needed and the extent of condition review that was assigned to the issue. In the condition report, the action initiated was, "Review 1RF19 VT

-2 activities to ensure scheduling and plant conditions meet ASME code requirements."

The inspectors found that this issue had been previously identified by the licensee during a self-assessment evaluation in 2009. The evaluation stated in part, "Making preparations for end of interval Class 1 pressure test per IWB

-5222(b) [Sub

-article of IWB-5000]. The pressure test may meet the normal pressure testing requirements, but may not be sufficient to meet the requirements of the end of interval pressure test on Class 1 systems." The issue was entered into the licensee's corrective action program as Condition Report CR

-2009-004987. The licensee incorrectly closed this issue with the determination that the system pressure tests met the ASME 10

-year pressure test requirements.

The only corrective action the licensee had assigned to the current condition report was to perform a review of the 1RF19 VT2 pressure tests. Based on the nature of the performance deficiency, interviews with the program owner as to the cause of the performance deficiency, as well as previous instances where NRC inspectors or the licensee had found required ASME Code examinations and tests had been missed or incorrectly performed, the inspectors determined that this action was too narrowly focused. The licensee entered this issue into the corrective action program and submitted a relief request to allow the use of ASME Code Cases N

-798 and N-800 to restore compliance with regulatory requirements. The NRC granted verbal Relief Request 1/2B 3-2 for the Unit 1 and Unit 2 Third Ten Year Inservice Inspection Interval on November 1 , 2017.

Analysis.

The inspectors determined that the licensee's failure to perform a system leakage test of all Class 1 pressure retaining components for each of the units was a performance deficiency. This finding was more than minor because it affected the Initiating Events Cornerstone attribute of equipment reliability, and affected the cornerstone objective to limit the likelihood of events that upset plant stability and challenge critical safety functions. In accordance with Inspection Manual Chapter 0609, Appendix A, "The Significance Determination Process (SDP) for Findings At

-Power," dated June 19, 2012, Exhibit 1, "Initiating Events Screening Questions,"

the finding was determined to be of very low safety significance (Green) because the finding did not result in exceeding the reactor coolant system leak rate for a small loss

-of-coolant accident, and did not affect other systems used to mitigate a loss

-of-coolant accident resulting in a total loss of their function. The finding has a problem identification and resolution cross

-cutting aspect associated with evaluation. Although the licensee identified that they had failed to properly perform pressure testing on portions of various ASME Code Class 1 systems, they failed to thoroughly evaluate the issue to ensure that their resolution fully addressed ASME requirements and the need to obtain a relief request prior to restarting Unit 1 following the outage [P.2].

Enforcement.

Title 10 CFR 50.55a(g)(4) requires that components classified as ASME Code Class 1, Class 2, and Class 3 meet the requirements set forth in Section XI of the applicable editions of the ASME Boiler and Pressure Vessel Code, and Addenda. Tit le 10 CFR 50.55a(g)(4)(ii) requires that inservice examination of components be conducted during successive 120

-month inspection intervals, and comply with the requirements of the latest edition and addenda of the Code applicable to the specific interval.

The AMSE Code,Section XI, Article IWB

-5000 requires, in part, that pressure retaining components shall be tested with a system leakage test conducted at a pressure not less than the pressure corresponding to 100 percent reactor power prior to startup at the end of each refueling outage, and at or near the end of each 10

-year interval. Contrary to these requirements, prior to November 1, 2017, the licensee failed to perform the required system leakage test at a pressure not less than the pressure corresponding to 100 percent reactor power on all Class 1 pressure retaining components for each of the units prior to startup at the end of each refueling outage, and at or near the end of each 10

-year interval. Each unit is currently in its third 10

-year interval. Because this finding is of very low safety significance

, a verbal relief request was granted , and this was entered into the corrective action program as Condition Reports CR

-2017-010530 and CR

-2017-011968, this violation is being treated as a NCV consistent with Section 2.3.2.a of the NRC Enforcement Policy.

(NCV 05000445/2017004

-0 2; 05000446/2017004

-0 2, Failure to Perform Class 1 Piping Pressure Tests in Accordance with ASME Code)1 R 11 Licensed Operator Requalification Program and Licensed Operator Performance (71111.11)

.1 Review of Licensed Operator Requalification

a. Inspection Scope

On November 21, 2017, the inspectors observed an evaluated simulator scenario performed by an operating crew. The inspectors assessed the performance of the operators and the evaluators' critique of their performance.

These activities constituted 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

I nspectors observed the performance of on

-shift licensed operators in the plant's main control room. At the time of the observations, the plant was in a period of heightened activity and risk. The inspectors observed the operators' performance of the following activities:

October 8, 2017, shutting down Unit 1 for refueling October 31, 2017, draining Unit 1 reactor vessel down to mid

-loop conditions In addition, the inspectors assessed the operators' adherence to plant procedures, including conduct of operations procedure and other operations department policies.

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

b. Findings

No findings were identified.

1 R 12 Maintenance Effectiveness (71111.12)

.1 Routine Maintenance Effectiveness

a. Inspection Scope

The inspectors reviewed four instances of degraded performance or condition of safety

-significant structures, systems, and components (SSCs):

November 17, 2017, Unit 1 containment spray system , full system review October 25, 2017, service water intake structure , delamination of concrete December 13, 2017, Units 1 and 2 Maintenance Rule (a)(3) assessment December 20, 2017, common instrument air system, compressor air leaks The inspectors reviewed the extent of condition of possible common cause SSC failures and evaluated the adequacy of the licensee's corrective actions. The inspectors reviewed the licensee's work practices to evaluate whether these may have played a role in the degradation of the SSCs. The inspectors assessed the licensee's 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 four maintenance effectiveness sample s, as defined in Inspection Procedure 71111.12.

b. Findings

Introduction.

The inspectors identified a Green non-cited violation of 10 CFR 50, Appendix B, Criterion XVI, "Corrective Action," associated with the licensee's failure to take timely corrective action s for a condition adverse to quality.

Specifically, the licensee failed to take corrective actions for indications of degraded concrete in the service water intake structure.

Description.

On October 17, 2017, while touring the service water intake structure pump bay, the inspectors noted concrete spalling resulting in large pieces of concrete falling onto the walkway. The inspectors also observed exposed and corroded rebar, and additional rust blooms in the overhead concrete. The inspectors found that the licensee had been aware of indications of concrete degradation since December 2010. The inspectors reviewed the structural monitoring program for the service water intake structure and determined that the licensee had documented indications of increasingly degraded concrete over many inspections, but had not taken corrective actions. The licensee had generated work orders to address the spalling, but did not complete the planned work. Prior to the inspectors' tour, the inspections had last been completed in June 2017.

The inspectors determined that a long term degraded condition had existed since at least 2010 resulting in corrosion of rebar in the concrete and ultimately resulting in significant spalling. The licensee determined that with the existing spalling sufficient concrete remained to maintain structural integrity. This issue does not represent an immediate safety concern because the licensee performed detailed engineering evaluations and tests to confirm the structural integrity of the intake structure, and will implement corrective actions to repair the affected concrete in the future. The licensee entered this issue into the corrective action program as Condition Report CR-2018-000088.

Analysis.

The licensee's failure to take timely and adequate corrective actions to correct a condition adverse to quality was a performance deficiency. The performance deficiency is more than minor , and therefore a finding, because it is 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 failure to correct the condition resulted in a degraded condition with no evaluation to determine continued operability. Using Inspection Manual Chapter 0609, Attachment 04, "Initial Characterization of Findings," dated October 7, 2016, and Inspection Manual Chapter 0609, Appendix A, "Significance Determination Process for Findings At

-Power," Exhibit 4 , "External Events Screening Questions," the inspectors determined the finding was of very low safety significance (Green) because the finding is a deficiency affecting the design or qualification of a mitigating SSC, but the SSC maintain ed its operability. The finding has a human performance cross

-cutting aspect associated with conservative bias, in that, the licensee failed to ensure that the individuals used decision making practices that emphasized prudent choices [H.14

].

Enforcement.

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, from December 2010 through October 2017, for quality related structures to which 10 CFR Part 50, Appendix B applies, the licensee failed to assure that conditions adverse to quality were promptly identified and corrected.

This issue does not represent an immediate safety concern because the licensee performed detailed engineering evaluations and tests and determined that the structural integrity of the intake structure remained adequate. Since this violation was of very low safety significance (Green) and has been entered into the corrective action program as Conditio n Report CR-2018-000088, this violation is being treated as a non

-cited violation consistent with Section 2.3.2.a of the NRC Enforcement Policy. (NCV 05000445/2017004

-03; 05000446/2017004

-03, Failure to Promptly Correct a Condition Adverse to Quality

) 1R 13 Maintenance Risk Assessments and Emergent Work Control (71111.13)

a. Inspection Scope

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

October 5, 2017, Unit 1 , risk mitigation plan during fuel movement October 18, 2017, Unit 1 and 2, opening control room door E

-40A November 22, 2017, Unit 2, diesel generator 2

-01 fiber optic cable replacement December 1, 2017, modification of XST1 with safety chiller 2

-05 out of service The inspectors verified that these risk assessment s 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 licensee's risk assessment s and verified that the licensee implemented appropriate risk management actions based on the result of the assessment s. Additionally, on November 8, 2017, the inspectors also observed portions of one emergent work activity that had the potential to affect the functional capability of mitigating systems. The inspectors observed emergent work on the Unit 1, train B safety related battery. 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 unaffected SSCs.

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

b. Findings

Introduction.

The inspectors identified a Green, non

-cited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, associated with the licensee's failure to accomplish activities affecting quality in accordance with documented procedures. Specifically, the licensee breached control room door E

-40A, a tornado missile barrier, without following the requirements of standing order OSO

-008.

Description.

Comanche Peak Inspection Report 2017002 document ed finding NCV 05000445/2017002

-04; 05000446/2017002

-04, Failure to Adequately Assess Risk and Implement Risk Management Actions for Proposed Maintenance. The cause of this finding was the licensee's failure to implement preplanned actions contained in station procedure ODA

-308, "LCO Tracking Program," Revision 16, section 13.7.39, "Tornado Missile Shields," when disabling a hazard barrier, door E

-40A. As a result of this finding the licensee developed and issued Standing Order OSO

-008, "Control of Control Room Door for Material Loadout," to proceduralize required actions when opening door E

-40A for other than routine passage. Specifically, operators were directed to implement the requirements of ODA

-308 section 3.7.10, "Control Room Emergency Filtration/Pressurization System," Section 13.7.39, "Tornado Missile Shields," and STA-759, "Control Room Envelop Habitability Program," Attachments 8.B and 8.C. Among other requirements, these references directed operators to station a dedicated operator in continuous communication at door E

-40A when opened for other than routine passage. While touring the control room on September 28, 2017, inspectors noted that doo r E-40A was open for maintenance activities with no one stationed at the door as required by Standing Order OSO

-008. Inspectors learned that the unit supervisor was unaware that the door had been opened, and that the clearance processing center senior reactor operator had authorized opening of the door. Furthermore inspectors determined that the clearance processing center senior reactor operator had failed to ensure that the requirements of Standing Order OSO

-008 were implemented prior to opening door E-40A. Subsequently, activities were stopped and door E

-40A was shut. Condition Report CR-2017-010661 was generated to capture this issue in the station's corrective action program.

Analyses. The licensee's failure to follow the requirements of Standing Order OSO

-008 when breaching door E

-40A was a performance deficiency. This performance deficiency is more than minor, and therefore a finding, because it is associated with protection against external factors 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.

Using Inspection Manual Chapter 0609, Attachment 04, "Initial Characterization of Findings," dated October 7, 2016, and Inspection Manual Chapter 0609, Appendix A, "Significance Determination Process for Findings At

-Power," Exhibit 2 , "Mitigating Systems Screening Questions," the inspectors determined the finding was of very low safety significance (Green) because:

(1) it was not a design deficiency;

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

(3) it did not represent an actual loss of function of at least a single train for longer than its technical specification allowed outage time;

(4) and it did not result in the loss of a high safety significant non

-technical specification train. The finding has a problem identification and resolution cross

-cutting aspect associated with resolution, in that the licensee failed to take adequate corrective actions from a previous issue with this door [P.3].

Enforcement.

Title 10 CFR Part 50, Appendix B, Criterion V, "Instructions, Procedures, and Drawings," requires, in part, that activities affecting quality shall be accomplished in accordance with documented instructions, procedures, or drawings, of a type appropriate to the circumstances. Standing Order OSO

-008, "Control of Control Room Door for Material Loadout," an Appendix B quality related procedure, provides instructions for breaching the control room boundary thru door E

-40A. Contrary to the above, on September 28, 2017, the licensee failed to follow the requirements of Standing Order OSO

-008 when breaching door E

-40A. Specifically, the licensee failed to implement the required compensatory measure prior to breaching a tornado missile barrier/control room envelop barrier. This issue does not represent an immediate safety concern because when this was identified the licensee took action to immediately close door E-40A. Since this violation was of very low safety significance (Green) and has been entered into the corrective action program as Condition Report CR

-2017-010661, this violation is being treated as a non

-cited violation consistent with Section 2.3.2.a of the NRC Enforcement Policy. (NCV 05000445/2017004

-04; 05000446/2017004

-04, Failure to Follow Procedure When Breaching Control Room Envelop e) 1 R 15 Operability Determinations and Functionality Assessments (71111.15)

a. Inspection Scope

The inspectors reviewed four operability determinations that the licensee performed for degraded or nonconforming SSCs: October 12, 2017, Unit 1, environmental qualification October 16, 2017, Unit 1 and 2, door maintenance on primary plant ventilation boundary door S1

-35G November 2, 2017, Unit 1, bent reach rod on safety related spring can for steam generator 1

-04 blowdown piping November 8, 2017, Unit 1, B safety related battery cell 41 jumpered out The inspectors reviewed the timeliness and technical adequacy of the licensee's evaluations. Where the licensee determined the degraded SSC to be operable, the inspectors verified that the licensee's compensatory measures were appropriate to provide reasonable assurance of operability. The inspectors verified that the licensee had considered the effect of other degraded conditions on the operability of the degraded SSC.

These activities constituted completion of four operability and functionality review sample s as defined in Inspection Procedure 71111.15.

b. Findings

Introduction

. Inspectors identified a Green , non-cited violation of 10 CFR 50, Appendix B, Criterion V, "Instructions, Procedures, and Drawings," associated with the licensee's failure to follow the requirements of station procedure STI

-606.01, "Work Control Process," Revision 5. Specifically, the work order for emergent maintenance activities on containment access corridor door S1

-35G was not reviewed by operations for plant impacts prior to work commencing. As a result, S1-35G was blocked open with the Unit 1 containment personnel airlock and equipment hatch open, causing the Unit 2 primary plant ventilation system to be inoperable.

Description.

On October 16, 2017, Unit 1 was shutdown in a refueling outage with the containment personnel airlock and equipment hatc h open. With Unit 1 in this configuration

, containment access corridor door S1

-35G is required to be shut so that the Unit 2 primary plant ventilation system boundary is maintained.

While touring the Unit 1 safeguards building

, inspectors noted that door S1-35G was blocked open for maintenance activities. Inspectors questioned this configuration because having previously been in the control room inspectors knew that Unit 2 had not entered Technical Specification 3.7.12, "Primary Plant Ventilation System," action statement A. Inspectors contacted the Unit 2 unit supervisor and informed the operations crew of the condition of door S1

-35G. The unit supervisor subsequently declared the primary plant ventilation system inoperable and entered the limiting condition for operation.

During subsequent review the licensee determined that Work Order 5508594 had not been reviewed by operations for plant impact. Specifically, station procedure STI

-606.01 Sections 5.16.1 and 6.8.7.1 requires, in part, that operations personnel review maintenance or work activities which impact plant equipment to determine effects on equipment operability. This resulted in the failure to identify the need to declare Unit 2 primary plant ventilation inoperable when the door was blocked open. Furthermore, the licensee also determined that inadequate communications between the work week manager and the Unit 2 unit supervisor regarding the scope of the work to be performed on door S1

-35G was a causal factor for this issue.

Analys is. The licensee's failure to follow procedural requirements and review work activities for plant impact when blocking open door S1

-35G for emergent maintenance was a performance deficiency. This performance deficiency is more than minor, and therefore a finding, because it is associated with the SSC and barrier performance attribute of the Barrier Integrity cornerstone and adversely affected the cornerstone objective to provide reasonable assurance that physical design barriers (fuel cladding, reactor coolant system, and containment) protect the public from radionuclide releases caused by accidents or events. Specifically, with the Unit 1 containment personnel airlock and equipment hatch open and door S1

-35G blocked open

, the Unit 2 primary plant ventilation system could not perform its function. Using Inspection Manual Chapter 0609, Attachment 04, "Initial Characterization of Findings," dated October 7, 2016, and Inspection Manual Chapter 0609, Appendix A, "Significance Determination Process for Findings At-Power," Exhibit 4, "External Events Screening Questions," the inspectors determined the finding to be of very low safety significance (Green) because it only represented a degradation of the radiological barrier function provided for the auxiliary building secondary containment. The finding has a human performance cross

-cutting aspect associated with team work, in that the licensee failed to ensure that work groups adequately communicate and coordinate their activities across organizational boundaries to ensure that nuclear safety is maintained

[H.4].

Enforcement.

10 CFR 50, Appendix B, Criterion V, "Instructions, Procedures, and Drawings," requires, in part, that activities affecting quality shall be accomplished in accordance with documented instructions, procedures, or drawings, of a type appropriate to the circumstances. Station procedure STI

-606.01, "Work Control Process," Revision 5, an Appendix B quality related procedure, provides instructions for planning maintenance activities on safety

-related equipment.

Contrary to the above, on October 16, 2017, an activity affecting quality was not accomplished in accordance with a procedure that was appropriate to the circumstances.

Specifically, procedure STI-606.01 , Sections 5.16.1 and 6.8.7.1 require, in part, that operations personnel review maintenance or work activities which impact plant equipment to determine effects on equipment operability.

This issue does not represent an immediate safety concern because the licensee took action to enter the required technical specification action statement and establish appropriate compensatory measures for the activity. Since this violation was of very low safety significance (Green) and has been entered into the corrective action program as Condition Report CR-2017-011424, this violation is being treated as a non

-cited violation consistent with Section 2.3.2.a of the NRC Enforcement Policy. (NCV 05000445/2017004

-05; 05000446/2017004

-05, Failure to Declare Equipment Inoperable)1 R 18 Plant Modifications (71111.18)

.1 Temporary Modifications

a. Inspection Scope

The inspectors reviewed two temporary plant modifications that affected risk

-significant SSCs: October 16, 2017, Unit 1, control room cabinets temporary modification for integrated system testing November 8, 2017, Unit 1, B train safety related battery cell 41 jumper The inspectors verified that the licensee had installed and removed these temporary modifications in accordance with technically adequate design documents. The inspectors verified that these modifications did not adversely impact the operability or availability of affected SSCs. The inspectors reviewed design documentation and plant procedures affected by the modifications to verify the licensee maintained configuration control. These activities constituted completion of two samples of temporary modifications, as defined in Inspection Procedure 71111.18.

b. Findings

No findings were identified.

1 R 19 Post-Maintenance Testing (71111.19)

a. Inspection Scope

The inspectors reviewed seven post-maintenance testing activities that affected risk

-significant SSCs:

September 26, 2017, Unit 1 , containment spray pump 1

-0 3 following maintenance October 25, 2017, Unit 2, steam generator 3 atmospheric relief valve following maintenance October 26, 2017, Unit 2, diesel generator 2

-01 following fiber optic cable replacement October 30, 2017, Unit 1, component cooling water discharge cross

-connect valve 1-HV-4515 following maintenance November 2, 2017, Unit 1, containment personnel airlock testing following limit switch adjustment November 17, 2017, Unit 1, auxiliary feedwater pump 1-01 following maintenance November 29, 2017, Unit 2, solid state protection system testing following replacement of switch S810 The inspectors reviewed licensin g- 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 constituted completion of seven post-maintenance testing inspection sample s, as defined in Inspection Procedure 71111.19.

b. Findings

No findings were identified.

1 R 20 Refueling and Other Outage Activities (71111.20)

a. Inspection Scope

During the station's Unit 1 refueling outage that concluded on November 6, 2017, the inspectors evaluated the licensee's 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:

revie w of the licensee's outage plan prior to the outage review and verification of the licensee's 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 constituted completion of one refueling outage sample as defined in Inspection Procedure 71111.20. On November 25, 2017, Unit 2 was manually tripped due to a simultaneous loss of both main feed water pumps, and the unit was not returned to full power until December 3, 2017. However, the unit shutdown and repair activities did not require the unit to cool down or a containment entry for a shutdown tour. Therefore

, the conditions were not met requiring the inspectors to formally complete the applicable sample elements defined in Inspection Procedure 71111.20 for the unplanned outage, though the inspectors did review the licensee's response to the unit trip in accordance with Inspection Procedure 71153 (see Section

4OA3 below).

b. Findings

Introduction.

The inspectors identified a Green , non-cited violation of 10 CFR 26.205 associated with the licensee's failure to calculate and control work hours of personnel subject to work hour controls

. Specifically, the licensee failed to track work hours for personnel assigned to emergency containment equipment hatch closure duty, resulting in multiple violations of work hour requirements.

Description.

During the Unit 1 refueling outage 1RF19, the inspectors reviewed the licensee work hour control program. The inspectors reviewed the work hour controls for personnel designated to perform rapid containment equipment hatch closure in response to abnormal events. The licensee uses an electronic database to track personnel work hours such that their work schedules do not violate the NRC's fatigue rule. The inspectors found that personnel designated for containment hatch closure and containment coordination were not included in the licensee's database, and subsequently raised this apparent discrepancy to licensee personnel.

The licensee reviewed procedure STA

-615, "Fatigue Management and Staff Work Hours," and determined that the personnel met the criteria for covered workers and should have been included in their program for controlling fatigue. The licensee concluded that 31 personnel had been incorrectly excluded from the program. Th e licensee determined that, during the Unit 1 refueling outage, 21 of these personnel exceeded fatigue rule work hour limits during the outage without an evaluation or waiver, resulting in 56 total violations of work hour limits. The inspectors reviewed logs and condition reports and determined that the personnel's fatigue did not result in any adverse consequences, and that emergency containment closure had not been required to be implemented during the outage.

The licensee entered this issue into the corrective action program as Condition Report CR-2017-012922. Through subsequent reviews the inspectors determined that the licensee relied on a manager's interpretation documented in a previous review of outage personnel subject to work hour limits conducted more than three years ago as their bases for not including these workers. Therefore inspectors determined that the performance deficiency was not reflective of current performance.

Analyses. The licensee's failure to control work hours for personnel subject to work hour controls was a performance deficiency. The performance deficiency is more than minor , and therefore a finding, because it is associated with the human performance attribute of the Barrier Integrity Cornerstone and affected the cornerstone objective to provide reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. Specifically, the failure to control work hours for personnel assigned to emergency containment hatch closure introduced a higher risk of human performance errors in maintaining containment integrity. Using Inspection Manual Chapter 0609, Attachment 04, "Initial Characterization of Findings," dated October 7, 2016, and Inspection Manual Chapter 0609, Appendix G, Attachment 1, "Shutdown Operations Significance Determination Process Phase 1 Initial Screening and Characterization of Findings," Exhibit 4, "Barrier Integrity Screening Questions," the inspectors determined the finding was of very low safety significance because it did not degrade the ability to close or isolate containment and did not degrade the physical integrity of reactor containment. The finding was not assigned a cross

-cutting aspect because the performance deficiency was not reflective of current performance.

Enforcement.

10 CFR 26.205 requires, in part, that licensees shall calculate and control the work hours of individuals who are subject to that section. Contrary to the above, from October 8, 2017 through November 1, 2017, for individuals performing duties to which 10 CFR 26 applies, the licensee failed to assure that work hours were calculated and controlled.

In response to this issue, the licensee developed actions to include the individuals assigned to emergency containment equipment hatch closure duty in work hour controls during future outages. This issue does not represent an immediate safety concern because at the time of discovery no personnel were in excess of work hour requirements, and the licensee took action to include the individuals in work hour controls during future outages. Since this violation was of very low safety significance (Green) and has been entered into the corrective action program as Condition Report CR-2017-012922, this violation is being treated as a non-cited violation consistent with Section 2.3.2.a of the NRC Enforcement Policy.

(NCV 05000445/2017004

-06; 05000446/2017004

-06, Failure to Control Work Hours for Covered Personnel)1 R 22 Surveillance Testing (71111.22)

a. Inspection Scope

The inspectors observ ed two 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:

Other surveillance tests:

October 30, 2017, Unit 1, component cooling water discharge cross

-connect valve 1-HV-4515 stroke testing October 26, 2017 , Unit 1 containment sump inspection 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 constituted completion of t wo surveillance testing inspection sample s, as defined in Inspection Procedure 71111.22.

b. Findings

No findings were identified.

RADIATION SAFETY

Cornerstone: Public Radiation Safety and Occupational Radiation Safety

2 RS 2 Occupational ALARA Planning and Controls (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 as a post

-outage review. 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 licensee's planning identified appropriate dose reduction techniqu es , reviewed any inconsistencies between intended and actual work activity doses , and determined if post-job (work activity) reviews were conducted to identify lessons learned.

Verification of dose estimates and exposure tracking systems, including the basis for exposure estimates, and measures to track, trend, and

, if necessary, reduce occupational doses for ongoing work activities. The inspectors evaluated the licensee's method for adjusting exposure estimates and reviewed the licensee's evaluations of inconsistent or incongruent results from the licensee's intended radiological outcomes.

Problem identification and resolution for ALARA planning. 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 three of the five required samples of occupational ALARA planning and controls program, as defined in Inspection Procedure 71124.02, and completes the inspection.

b. Findings

No findings were identified.

2 RS 4 Occupational Dose Assessment (71124.04)

a. Inspection Scope

The inspectors evaluated the accuracy and operability of the licensee's personnel monitoring equipment, verified the accuracy and effectiveness of the licensee's methods for determining total effective dose equivalent, and verified that the licensee was appropriately monitoring occupational dose.

The inspectors interviewed licensee personnel, walked down various portions of the plant, and reviewed licensee performance in the following areas:

source term characterization, including characterization of radiation types and energies, hard to detect isotopes, and scaling factors external dosimetry including National Voluntary Laboratory Accreditation Program (NVLAP) accreditation , storage, issue, use, and processing of active and passive dosimeters internal dosimetry, including the licensee's use of whole body counting, use of in vitro bioassay methods, dose assessments based on airborne monitoring, and the adequacy of internal dose assessments special dosimetric situations, including dec lared pregnant workers, dosimeter placement and assessment of effective dose equivalent for external exposures (EDEX), shallow dose equivalent , and neutron dose assessment problem identification and resolution for occupational dose assessment, including review of audits, self

-assessments, and corrective action program documents to verify problems were being identified and properly addressed for resolutio n These activities constitute completion of the five required samples of occupational dose assessment program, as defined in Inspection Procedure 71124.04.

b. Findings

No findings were identified.

OTHER ACTIVITIES

Cornerstones: Initiating Events, Mitigating Systems, Barrier Integrity, Emergency Preparedness, Public Radiation Safety, Occupational Radiation Safety, and Security 4OA 1 Performance Indicator Verification (71151)

.1 Mitigating Systems Performance Index: Residual Heat Removal Systems (MS09

)

a. Inspection Scope

The inspectors reviewed the licensee's mitigating system performance index data for the period of October 1, 2016, through September 30, 2017, to verify the accuracy and completeness of the reported data. 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 constituted verification of the mitigating system performance index for residual heat removal systems for Units 1 and 2, as defined in Inspection Procedure 71151.

b. Findings

No findings were identified.

.2 Mitigating Systems Performance Index: Cooling Water Support Systems (MS10)

a. Inspection Scope

The inspectors reviewed the licensee's mitigating system performance index data for the period of October 1, 2016, through September 30, 2017, to verify the accuracy and completeness of the reported data. 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 constituted verification of the mitigating system performance index for cooling water support systems for Units 1 and 2, as defined in Inspection Procedure 71151.

b. Findings

No findings were identified.

4OA 2 Problem Identification and Resolution (71152)

.1 Routine Review

a. Inspection Scope

Throughout the inspection period, the inspectors performed daily reviews of items entered into the licensee's corrective action program and periodically attended the licensee's condition report screening meetings. The inspectors verified that licensee personnel were identifying problems at an appropriate threshold and entering these problems into the corrective action program for resolution. The inspectors verified that the licensee developed and implemented corrective actions commensurate with the significance of the problems identified. The inspectors also reviewed the licensee's problem identification and resolution activities during the performance of the other inspection activities documented in this report.

b. Findings

No findings were identified.

.2 Annual Follow

-up of Selected Issues

a. Inspection Scope

The inspectors selected one issue for an in-depth follow

-up: The inspectors reviewed the licensee's corrective action program and associated documents to identify trends that could indicate the existence of a more significant safety issue.

The inspectors focused on an issue associated with incorrect shaft sizing for the turbine driven auxiliary feedwater pumps documented in Condition Report CR

-2017-004821. The inspectors assessed the licensee's 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 constituted completion of one annual follow

-up sample as defined in Inspection Procedure 71152.

b. Findings

No findings were identified.

4OA 3 Follow-up of Events and Notices of Enforcement Discretion (71153)

.1 Manual Reactor Trip Following A Loss of Main Feedwater

a. Inspection Scope

On November 25, 2017, during normal operations at 100 percent reactor power, the on shift operators of Comanche Peak Unit 2 inserted a manual reactor trip following a loss of all main feedwater due to a simultaneous loss of both main feedwater pumps. The plant responded with no complications. The operators placed the unit in Mode 3 with steam dumps and the auxiliary feedwater system in operation.

The inspectors evaluated operator performance, and reviewed operator logs, and plant computer data. Inspectors also reviewed the licensee's evaluation of the system response and cause of the trip, as well as the licensee's immediate corrective actions.

Inspectors did not identify a performance deficiency during the initial plant response to the event.

b. Findings

No findings were identified.

These activities constituted completion of one event follow

-up sample, as defined in Inspection Procedure 71153. 4OA 6 Meetings, Including Exit

Exit Meeting Summary

On October 25, 2017, the inspectors presented the inservice inspection results to Mr. T. McCool, Site Vice President, and other members of the licensee staff. The inspectors re

-exited the se inspection results to Mr. Jack Hicks, Regulatory Affairs, and other members of the licensee staff on December 11 , 2017. The licensee acknowledged the issues presented. The licensee confirmed that any proprietary information reviewed by the inspectors had been returned or destroyed.

On December 7, 2017, the inspectors presented the radiation safety inspection results to Mr. K. Peters, Senior Vice President and Chief Nuclear Officer, 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.

On January 3, 2018, the inspectors presented the overall resident inspection results to Mr. Ken Peters, Senior Vice President and Chief Nuclear Officer, 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.

SUPPLEMENTAL INFORMATION

KEY POINTS OF CONTACT

Licensee Personnel

K. Peters, Senior Vice President and Chief Nuclear Officer

T. McCool, Site Vice President

S. Dixon, Consulting License Analyst, Regulatory Affairs

J. Dreyfuss, Manager, Plant

R. Garcia, Supervisor, Radiation Protection.

J. Goodrich, Supervisor, Radiation Protection.

J. Gumnick, Manager, Radiation Protection

R. Knapp, Supervisor, Radiation Protection

T. Hope, Manager, Regulatory Affairs

K. Mandrell, Supervisor, Radiation Protection.

E. McGurk, Supervisor, Radiation Protection

G. Merka, License Analyst, Regulatory Affairs

L. Windham, Manager, Corrective Action Program

D. Davis, Manager, Organizational Development

R. Deppi, Director, Project Engineering and Support

D. Goodwin, Director, Work Management

T. Hope, Manager, Regulatory Affairs

J. Hull, Manager, Nuclear Emergency Preparedness

A. Marzloft, Director, Nuclear Oversight

K. Peters, Senior Vice President and Chief Nuclear Officer

D. Volkening, Audit Manager, Nuclear Oversight

S. Sewell, Director, Engineering and Regulatory Affairs

J. Taylor, Director, Site Engineering

C. Tran, Manager, Engineering Programs

LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED

Opened and Closed

05000445/2017004

-01

05000446/2017004

-01 NCV Inadequate Procedure for Containment Closure

05000445/2017004

-0 2

05000446/2017004

-0 2 NCV Failure to

Perform Class 1 Piping Pressure Tests in Accordance with ASME Code

05000445/2017004

-0 3

05000446/2017004

-0 3 NCV Failure to Promptly Correct a Condition Adverse to Quality

05000445/2017004

-0 4

05000446/2017004

-0 4 NCV Failure to Follow Procedure When Breaching

Control Room Envelop e

05000445/2017004

-0 5

05000446/2017004

-0 5 NCV Failure to Declare Equipment Inoperable

05000445/2017004

-0 6

05000446/2017004

-0 6 NCV Failure to Control Work Hours for Covered Personnel