IR 05000266/2019003

Integrated Inspection Report 05000266/2019003 and 05000301/2019003

ML19305D854
Person / Time
Site:	Point Beach
Issue date:	11/01/2019
From:	Eric Duncan Reactor Projects Region 3 Branch 4
To:	Craven R Point Beach
References
IR 2019003
Download: ML19305D854 (25)

Text

ber 1, 2019

SUBJECT:

POINT BEACH NUCLEAR PLANT - INTEGRATED INSPECTION REPORT 05000266/2019003 AND 05000301/2019003

Dear Mr. Craven:

On September 30, 2019, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at Point Beach Nuclear Plant and 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.

One finding of very low safety significance (Green) is documented in this report. This finding involved a violation of NRC requirements. We are treating this violation as a non-cited violation (NCV) consistent with Section 2.3.2 of the Enforcement Policy.

A licensee-identified violation which was determined to be of very low safety significance is documented in this report. We are treating this violation as a non-cited violation (NCV)

consistent with Section 2.3.2 of the Enforcement Policy.

If you contest the violation or significance or severity of the violation documented in this inspection report, you should provide a response within 30 days of the date of this inspection report, with the basis for your denial, to the U.S. Nuclear Regulatory Commission, ATTN:

Document Control Desk, Washington, DC 20555-0001; with copies to the Regional Administrator, Region III; the Director, Office of Enforcement; and the NRC Resident Inspector at Point Beach.

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 III; and the NRC Resident Inspector at Point Beach. This letter, its enclosure, and your response (if any) will be made available for public inspection and copying at http://www.nrc.gov/reading-rm/adams.html and at the NRC Public Document Room in accordance with 10 CFR 2.390, Public Inspections, Exemptions, Requests for Withholding.

Sincerely,

/RA/

Eric R. Duncan, Chief Branch 4 Division of Reactor Projects Docket Nos. 05000266 and 05000301 License Nos. DPR-24 and DPR-27

Enclosure:

As stated

Inspection Report

Docket Numbers: 05000266 and 05000301 License Numbers: DPR-24 and DPR-27 Report Numbers: 05000266/2019003 and 05000301/2019003 Enterprise Identifier: I-2019-003-0060 Licensee: NextEra Energy Point Beach, LLC Facility: Point Beach Nuclear Plant Location: Two Rivers, WI Inspection Dates: July 01, 2019 to September 30, 2019 Inspectors: K. Barclay, Resident Inspector G. Edwards, Health Physicist T. Hartman, Senior Resident Inspector J. Neurauter, Senior Reactor Inspector D. Szwarc, Senior Reactor Inspector Approved By: Eric R. Duncan, Chief Branch 4 Division of Reactor Projects Enclosure

SUMMARY

The U.S. Nuclear Regulatory Commission (NRC) continued monitoring the licensees performance by conducting an integrated inspection at Point Beach Nuclear Plant in accordance with the Reactor Oversight Process. The Reactor Oversight Process is the NRCs program for overseeing the safe operation of commercial nuclear power reactors. Refer to https://www.nrc.gov/reactors/operating/oversight.html for more information. A licensee-identified non-cited violation is documented in report section: 71152.

List of Findings and Violations Failure to Barricade and Conspicuously Post a High Radiation Area in Unit 1 Containment Cornerstone Significance Cross-Cutting Report Aspect Section Occupational Green [H.4] - 71124.01 Radiation Safety NCV 05000266,05000301/2019003-01 Teamwork Open/Closed A self-revealed finding of very low safety significance (i.e., Green) and an associated Non-Cited Violation (NCV) of Technical Specification 5.7.1(a), High Radiation Areas, was identified by the inspectors when licensee personnel failed to barricade and conspicuously post a high radiation area on the 8 foot elevation of the Unit 1 containment. As a result, two workers inadvertently entered a high radiation area with dose rates greater than 100 millirem (mrem)/hour.

Additional Tracking Items Type Issue Number Title Report Section Status LER 05000266,05000301/20 LER 2019-002-00 for Point 71153 Closed 19-002-00 Beach Nuclear Plant, Units 1 and 2, Operation Not Maintained within Pressure and Temperature Limit Report (PTLR) as Required by Technical Specification 3.4.3 LER 05000266/2019-001-00 LER 2019-001-00 for Point 71153 Closed Beach Nuclear Plant, Unit 1,

Loss of Main Condenser Cooling Results in Manual Reactor Trip

PLANT STATUS

Unit 1 operated at or near rated thermal power for the entire inspection period.

Unit 2 began the inspection period at rated thermal power. On August 12, 2019, power was reduced to about 50 percent after a control rod inadvertently fully inserted into the core. On August 14, 2019, power was reduced to about 40 percent for recovery of the control rod. Unit 2 returned to rated thermal power on August 16, 2019, and remained at or near rated thermal power for the remainder of the inspection period.

INSPECTION SCOPES

Inspections were conducted using the appropriate portions of the inspection procedures (IPs) in effect at the beginning of the inspection unless otherwise noted. Currently approved IPs with their attached revision histories are located on the public website at http://www.nrc.gov/reading-rm/doc-collections/insp-manual/inspection-procedure/index.html. Samples were declared complete when the IP requirements most appropriate to the inspection activity were met consistent with Inspection Manual Chapter (IMC) 2515, Light-Water Reactor Inspection Program - Operations Phase. The inspectors performed plant status activities described in IMC 2515 Appendix D, Plant Status and conducted routine reviews using IP 71152, Problem Identification and Resolution. The inspectors reviewed selected procedures and records, observed activities, and interviewed personnel to assess licensee performance and compliance with Commission rules and regulations, license conditions, site procedures, and standards.

71111.01 - Adverse Weather Protection Seasonal Extreme Weather Sample (IP Section 03.02)

===(1) The inspectors evaluated the licensee's readiness for seasonal extreme weather conditions prior to the onset of seasonal high temperatures for the following systems:

• Emergency Diesel Generators (EDGs)

• 4160 Volt Electrical Distribution

• 120 Volt Alternating Current (AC) Vital Instrument Power

• 125 Volt Direct Current (DC) Electrical Distribution

71111.04Q - Equipment Alignment Partial Walkdown Sample (IP Section 03.01)

The inspectors evaluated system configurations during partial walkdowns of the following systems/trains:

(1) Service Water Ring Header Continuous Flowpath Interrupted with Alternate Alignment implemented on July 25, 2019

(2) Unit 2 Turbine-Driven Auxiliary Feedwater Pump (TDAFP) following testing on September 7, 2019

(3) G-01 Emergency Diesel Generator (EDG) following testing on September 24, 2019

71111.05Q - Fire Protection Quarterly Inspection (IP Section 03.01)

The inspectors evaluated fire protection program implementation in the following selected areas:

(1) Fire Zone 301 on August 16, 2019

(2) Fire Zones 319 and 320 on August 26, 2019

(3) Fire Zones 300 and 303 on August 26, 2019

(4) Fire Zone 380 on August 26, 2019

(5) Fire Zone 318 on September 17, 2019

71111.06 - Flood Protection Measures Inspection Activities - Underground Cables (IP Section 02.02c.)

The inspectors evaluated cable submergence protection in:

(1) Cable Vaults Z-065A-H, J, P, Q-V; Z-066A-D; Z-067A-D; and Z-068

71111.11Q - Licensed Operator Requalification Program and Licensed Operator Performance Licensed Operator Performance in the Actual Plant/Main Control Room (IP Section 03.01)

(1) The inspectors observed and evaluated licensed operator performance in the Control Room during plant power reduction, rod recovery, and power ascension following a Unit 2 dropped control rod event from August 12 through August 15, 2019 Licensed Operator Requalification Training/Examinations (IP Section 03.02)===

(1) The inspectors observed and evaluated a licensed operator simulator scenario on August 5, 2019

71111.12 - Maintenance Effectiveness Routine Maintenance Effectiveness Inspection (IP Section 02.01)

===The inspectors evaluated the effectiveness of routine maintenance activities associated with the following equipment and/or safety significant functions:

(1) Periodic evaluation of 10 CFR 50.65(a)(3)

71111.13 - Maintenance Risk Assessments and Emergent Work Control Risk Assessment and Management Sample (IP Section 03.01)

The inspectors evaluated the risk assessments for the following planned and emergent work activities:

(1) Elevated risk during IT 45, Safety Injection Train B, on July 11, 2019

(2) Elevated risk during Bus A-03 protective relay calibration, on July 18, 2019

(3) Elevated risk with the South Service Water Header Strainer out-of-service and ring header interrupted from July 23-25, 2019

(4) Elevated risk and emergent work following a Unit 2 dropped Control Rod F-6 event, on August 12, 2019

(5) Elevated risk during the G-05 Gas Turbine maintenance window from September 8-14, 2019

71111.15 - Operability Determinations and Functionality Assessments Operability Determination or Functionality Assessment (IP Section 02.02)

The inspectors evaluated the following operability determinations and functionality assessments:

(1) Clarification Required Regarding Implementation of Limiting Condition for Operation (LCO) 3.4.3 on April 11, 2019

(2) G-02 EDG Step Change in G2H Vibration Readings on July 14, 2019

(3) G-04 EDG Exhaust Pipe Support Degraded on September 11, 2019

(4) G-03 EDG Sliding Feet Bolt and Hole Binding Indications on September 30, 2019

71111.18 - Plant Modifications Temporary Modifications and/or Permanent Modifications (IP Section 03.01 and/or 03.02)

The inspectors evaluated the following temporary or permanent modifications:

(1) Engineering Change 292242, Contingent Supports for Unit 1 Steam Generator/Reactor Coolant Pump A & B Platforms, on September 26, 2019

71111.19 - Post-Maintenance Testing Post-Maintenance Test Sample (IP Section 03.01)

The inspectors evaluated the following post maintenance tests:

(1) SW 1912, Service Water Header South Strainer Test, following maintenance, on July 25, 2019

(2) 0-PT-FP-002, Diesel Engine-Driven Fire Pump Functional Test, following maintenance, on July 30, 2019

(3) 1P-11A, Component Cooling Water Pump Test, following maintenance, on August 20, 2019

(4) 1P-2C, Charging Pump Test, following maintenance, on August 28-29, 2019 and September 11, 2019

(5) G-05, Gas Turbine Generator Test, following maintenance, on September 12-13, 2019

(6) G-03, EDG Test, following maintenance, on September 24, 2019

==71111.22 - Surveillance Testing The inspectors evaluated the following surveillance tests:

Surveillance Tests (other) (IP Section 03.01)==

(1) 1ICP 02.032, 1P-29 Auxiliary Feedwater Suction Header Pressure Trip Channel Operability Test, on July 25, 2019

(2) IT 07G, Service Water Valves Quarterly, on August 21, 2019

(3) 2ICP 02.003B, Reactor Protection System Logic Train B 31 Day Surveillance Test, on August 22, 2019 Inservice Testing (IP Section 03.01)===

(1) IT 12 Train A, 1P-11A Component Cooling Water Pump and Valves Unit 1, on

August 20, 2019 FLEX Testing (IP Section 03.02) (1 Sample)

(1) Z-2004A FLEX Portable Diesel Steam Generator Injection Pump Full Flow Testing, on September 4, 2019

71114.06 - Drill Evaluation Drill/Training Evolution Observation (IP Section 03.02)

The inspectors evaluated:

===(1) A crew simulator evaluation with Drill Exercise Performance (DEP), on August 5,

RADIATION SAFETY

==71124.01 - Radiological Hazard Assessment and Exposure Controls High Radiation Area and Very High Radiation Area Controls (IP Section 02.05) (1 Partial)

(1) (Partial)

The inspectors evaluated high radiation area controls, including a high radiation area that was created on the 8 foot elevation of the Unit 1 containment building.

71124.02 - Occupational ALARA Planning and Controls Radiological Work Planning (IP Section 02.01)==

The inspectors evaluated the licensees radiological work planning.

(1) The inspectors reviewed the following activities:

• 19-1016; Removal/Reinstall Reactor Vessel Head

• 19-1053; Unit 1 Narrow Range Resistance Temperature Detector (RTD)

Replacement

• 19-1057; Steam Generator Eddy Current Testing Verification of Dose Estimates and Exposure Tracking Systems (IP Section 02.02)===

The inspectors evaluated dose estimates and exposure tracking.

(1) The inspectors reviewed the following as-low-as reasonably-achievable (ALARA)planning documents:

• 19-1016; Removal/Reinstall Reactor Vessel Head

• 19-1053; Unit 1 Narrow Range Resistance Temperature Detector (RTD)

Replacement

• 19-1057; Steam Generator Eddy Current Testing

• 19-1064; Unit 1 B Reactor Coolant Pump Impeller Change Out and Pump Bowl Inspection Additionally, the inspectors reviewed the following radiological outcome evaluations:

• 19-1021; Reactor Coolant Pump Work; Post Outage Job Review

• 19-1064; Unit 1 B Reactor Coolant Pump Impeller Change Out and Pump Bowl Inspection; Post Outage Job Review

• 19-1057; Steam Generator Eddy Current Testing; Post Outage Job Review

71124.03 - In-Plant Airborne Radioactivity Control and Mitigation Self-Contained Breathing Apparatus for Emergency Use (IP Section 02.03)

===The inspectors evaluated self-contained breathing apparatus (SCBA) program implementation.

(1) The inspectors reviewed the following:

Status and Surveillance Records for Self-Contained Breathing Apparatus

• SCBA-K-47518

• SCBA-K-47977

• SCBA-K-48560 Self-Contained Breathing Apparatus Fit for On-Shift Operators

• Day On-Shift Operator 1

• Day On-Shift Operator 2

• Day On-Shift Operator 3 Self-Contained Breathing Apparatus Maintenance Check

• SCBA-K-12952

• SCBA-K-12709

• SCBA-K-12944

OTHER ACTIVITIES - BASELINE

===71151 - Performance Indicator Verification The inspectors verified licensee performance indicators submittals listed below:

MS06: Emergency AC Power Systems (IP Section 02.05)===

(1) Unit 1 July 1, 2018 - June 30, 2019

(2) Unit 2 July 1, 2018 - June 30, 2019 MS07: High Pressure Injection Systems (IP Section 02.06) ===

(1) Unit 1 July 1, 2018 - June 30, 2019

(2) Unit 2 July 1, 2018 - June 30, 2019

MS08: Heat Removal Systems (IP Section 02.07) (2 Samples)

(1) Unit 1 July 1, 2018 - June 30, 2019

(2) Unit 2 July 1, 2018 - June 30, 2019 BI01: Reactor Coolant System (RCS) Specific Activity Sample (IP Section 02.10) (2 Samples)

(1) Unit 1 July 1, 2018 - August 30, 2019

(2) Unit 2 July 1, 2018 - August 30, 2019 OR01: Occupational Exposure Control Effectiveness Sample (IP Section 02.15) (1 Sample)

(1) July 1, 2018 - August 30, 2019 PR01: Radiological Effluent Technical Specifications/Offsite Dose Calculation Manual Radiological Effluent Occurrences (RETS/ODCM) Radiological Effluent Occurrences Sample.

(IP Section 02.16) (1 Sample)

(1) July 1, 2018 - August 30, 2019

71152 - Problem Identification and Resolution Annual Follow-up of Selected Issues (IP Section 02.03)

===The inspectors reviewed the licensees implementation of its corrective action program related to the following issues:

(1) Thermal Loading from a Design-Basis Accident (DBA) Exceeds Platform Deflection Criteria

(2) Multiple Blown Fuses on Unit 2 Control Rod F-6

(3) Fire-Induced Short Affecting Battery Chargers D-09 and D-109

71153 - Follow-up of Events and Notices of Enforcement Discretion Event Report (IP Section 03.02)

The inspectors evaluated the following licensee event reports (LERs):

(1) LER 2019-001-00, Loss of Main Condenser Cooling Results in Manual Reactor Trip (ADAMS Accession ML19035A127). The inspectors determined that there was no regulatory requirement or self-imposed licensee standard that was not met.

Therefore, no performance deficiency was identified. The inspectors also concluded that no violation of NRC requirements occurred.

(2) LER 2019-002-00, Operation Not Maintained Within Pressure and Temperature Limit Report (PTLR) as Required by Technical Specification 3.4.3 (ADAMS Accession ML19161A083). The circumstances surrounding this LER are documented in Inspection Results Section 71153 of this report.

Personnel Performance (IP Section 03.03) ===

(1) The inspectors evaluated a plant transient and assessed the licensees response following a Unit 2 dropped control rod event on August 12,

INSPECTION RESULTS

Failure to Barricade and Conspicuously Post a High Radiation Area in Unit 1 Containment Cornerstone Significance Cross-Cutting Report Aspect Section Occupational Green [H.4] - 71124.01 Radiation Safety NCV 05000266,05000301/2019003-01 Teamwork Open/Closed A self-revealed finding of very low safety significance (i.e., Green) and an associated Non-Cited Violation (NCV) of Technical Specification 5.7.1(a), High Radiation Areas, was identified by the inspectors when licensee personnel failed to barricade and conspicuously post a high radiation area on the 8 foot elevation of the Unit 1 containment. As a result, two workers inadvertently entered a high radiation area with dose rates greater than 100 millirem (mrem)/hour.

Description:

On April 12, 2019, Radiation Protection (RP) staff performed a cavity decontamination in the Unit 1 containment. Prior to RP performing this task, a briefing was conducted by RP with Operations staff, which included all activities associated with the cavity decontamination.

Some of these activities included the removal of the tri-nuke filter system and the flushing of the cavity drain line. During the briefing, RP informed Operations that RP would not be able to support the flushing of the cavity drain line at that time due to RP resources being needed in other areas of the plant, which meant that the flushing of the cavity drain line would have to be rescheduled.

The Operations staff recognized that the flushing would need to be delayed but believed that it would be acceptable to take some actions to prepare for the flushing. When the briefing concluded, Operations staff performed the valve line-up associated with the flushing of the cavity drain line. Although this valve line-up was not the task of flushing the cavity drain line, the valve line-up caused water within the system to shift within the cavity drain line. When this water shift occurred, a hot spot of elevated radioactivity was formed within the cavity drain line on the 8 foot elevation of the Unit 1 containment.

On April 12, 2019, two Instrument and Control (I&C) Technicians entered the general area associated with the hot spot and received electronic dosimeter dose rate alarms, that were set at 80 mrem/hour. Upon receiving these dose rate alarms the individuals immediately exited the area and reported the event to RP staff. The electronic dosimeters recorded dose rates of 131 mrem/hour and 109 mrem/hour. The individuals received no appreciable recorded dose associated with these dose rate alarms. The RP staff conducted surveys to identify the source of these dose rate alarms. The surveys revealed the exact location of the elevated radioactivity hot spot. The surveys indicated that the dose rates near the hot spot were 27.8 rem/hour on contact and 9 rem/hour at 30 centimeters. This elevated dose rate area was located approximately 12 feet from the ground level in the area. The inspectors determined that this area was not accessible. However, dose rates that were accessible to workers were approximately 300 mrem/hour. The RP staff then barricaded and posted the area as a high radiation area. The area had previously been posted as a radiation area.

Corrective Actions: The RP staff investigated and discovered that Operations staff had performed a system valve alignment of the cavity drain line without contacting RP. The licensee subsequently revised licensee procedure, RP-1D Filling and Draining the Unit 1 Refueling Cavity and Unit 1 Cavity Purification for Operations, to require that RP staff be contacted prior to conducting system alignments.

Corrective Action References: AR 02310391

Performance Assessment:

Performance Deficiency: The licensee failed to barricade and conspicuously post a high radiation area on the 8 foot elevation of Unit 1 containment. As a result, two workers inadvertently entered a high radiation area with dose rates greater than 100 mrem/hour.

Screening: The inspectors determined the performance deficiency was more than minor because it was associated with the Program & Process attribute of the Occupational Radiation Safety cornerstone and adversely affected the cornerstone objective to ensure the adequate protection of the worker health and safety from exposure to radiation from radioactive material during routine civilian nuclear reactor operation. Specifically, the failure to barricade and conspicuously post this high radiation area resulted in two workers inadvertently entering a high radiation area with uncharacterized radiological conditions that exceeded 100 mrem/hour.

Significance: The inspectors assessed the significance of the finding using Appendix C, Occupational Radiation Safety SDP. The inspectors determined that the finding was of very low safety significance (i.e., Green) because:

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

(2) there was no overexposure,

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

(4) the ability to assess dose was not compromised.

Cross-Cutting Aspect: H.4 - Teamwork: Individuals and work groups communicate and coordinate their activities within and across organizational boundaries to ensure nuclear safety is maintained. Specifically, the licensee determined that Operations staff performed a system valve alignment associated with the cavity drain line flush evolution without informing RP staff. The RP staff informed Operations that RP would be unable to support activities associated with the cavity drain line flush. Operations staff assumed that the valve line-up could be performed without RP support, which led to an unbarricaded and unposted high radiation area.

Enforcement:

Violation: Title 10 CFR 20.1901(b), Posting of High Radiation Areas, requires, in part, that the licensee shall post each high radiation area with a conspicuous sign or signs bearing the radiation symbol and the words, CAUTION, HIGH RADIATION AREA, or, DANGER, HIGH RADIATION AREA.

Technical Specification 5.7.1(a), High Radiation Areas, requires, in part, that each entryway to areas with dose rates greater than 100 mrem/hour at 1 foot and not exceeding 1 rem/hour at 30 centimeters from the radiation source, or any surface penetrated by the radiation shall be barricaded and conspicuously posted as a high radiation area and that such barricades shall be opened as necessary to permit entry or exit of personnel or equipment.

Contrary to the above, on April 12, 2019, the licensee failed to barricade and to conspicuously post entryways to a high radiation area on the 8 foot elevation of the Unit 1 containment as a high radiation area with a conspicuous sign or signs bearing the radiation symbol and the words CAUTION, HIGH RADIATION AREA or DANGER, HIGH RADIATION AREA, as required by 10 CFR 20.1901(b) and TS 5.7.1(a). Specifically, an area in the 8 foot elevation of the Unit 1 containment with radiation levels of about 300 mrem/hour at 1 foot, which was greater than 100 mrem/hour at 1 foot and did not exceed 1 rem/hour at 30 centimeters from the radiation source, or any surface penetrated by the radiation was not barricaded or posted as a high radiation area as required.

Enforcement Action: This violation is being treated as an non-cited violation, consistent with Section 2.3.2 of the Enforcement Policy.

Licensee-Identified Non-Cited Violation 71152 This violation of very low safety significance was identified by the licensee and has been entered into the licensee corrective action program and is being treated as a non-cited violation, consistent with Section 2.3.2 of the Enforcement Policy.

Violation: Point Beach License Condition 4.F requires, in part, that the licensee implement and maintain in effect all provisions of the approved Fire Protection Program that complies with 10 CFR 50.48(a) and 10 CFR 50.48(c), NFPA [National Fire Protection Association]

Standard NFPA 805, as approved in the Safety Evaluation Report (SER) dated September 8, 2016.

Section 2.4.2.1 of NFPA 805 states that, A comprehensive list of systems and equipment and their interrelationships to be analyzed for a fire event shall be developed. The equipment list shall contain an inventory of those critical components required to achieve the nuclear safety performance criteria of Section 1.5. Components required to achieve and maintain the nuclear safety functions and components whose fire-induced failure could prevent the operation or result in the maloperation of those components needed to meet the nuclear safety criteria shall be included. Availability and reliability of equipment selected shall be evaluated.

Contrary to the above, from February 14, 2017, until August 30, 2019, the licensee failed to implement and maintain in effect all provisions of the approved Fire Protection Program that complies with 10 CFR 50.48(a) and 10 CFR 50.48(c). Specifically, the licensee failed to evaluate 22 interrelationships between components required to achieve and maintain the nuclear safety functions whose fire-induced failure could prevent the operation or result in the maloperation of those components needed to meet the nuclear safety criteria of Section 1.5 as specified in Section 2.4.2.1 of NFPA 805.

Significance/Severity: Green. A regional Senior Reactor Analyst reviewed the licensee's risk analysis and confirmed through walkdowns and independent verifications that the delta-CDF

[Core Damage Frequency] results were less than 1E-6/year, and therefore the performance deficiency associated with this NCV was of very low safety significance (i.e., Green).

Corrective Action References: AR 2302088 Observation: Potential for Fire-Induced Cable Failures 71152 The inspectors performed a detailed review of Condition Report (AR) 2302088, Failure Mode Not Properly Considered for Battery Chargers. The licensee documented in the AR that their circuits analysis failed to address the potential for fire-induced cable failures for battery chargers D-09 and D-109, which resulted in the potential to inadvertently cross-tie energized 480 Volt (V) buses, which could result in the loss of these buses and damage to plant equipment. This potential failure was not analyzed by the licensee as required by Section 2.4.2.1 of NFPA 805.

Following identification, the licensee entered this issue into their corrective action program as AR 2302088. The licensee subsequently performed evaluation P3257-1000-00, Independent Evaluation of AR 02302088, Revision 0, which reviewed the concerns identified in AR 2302088. The licensee analyzed fire-induced circuit failure modes associated with cables ZF1491C, ZE2391C, ZE13212HC and ZF24212BC, considered inter-cable and intra-cable hot shorts, and the potential failure of both EDG trains. The inspectors reviewed AR 2302088 and evaluation P3257-1000-00 and concluded that the licensee had reasonably evaluated the additional fire-induced circuit failure modes. Following the inspectors review, the licensee identified and documented an additional 21 potential cross-tie connections that could result in bus damage in report PBN-BFJR-19-033, PBN Risk Impact of Spurious Breaker Operation Due to Fire, Revision 0.

The inspectors reviewed the licensees extent of condition evaluation report PBN-BFJR-19-033 and did not identify any additional concerns.

A licensee-identified violation is documented in Section 71152 of this report.

Observation: Reactor Coolant System Loop Platform Thermal Loading Concerns 71152 The inspectors performed a review of issues entered into the licensees corrective action program associated with platforms within the steam generator and reactor coolant pump cubicles. The inspectors noted that in AR 2286174 the licensee documented that a preliminary analysis concluded that during a design basis accident thermal loading would exceed the structural capacity of two of the four steam generator and reactor coolant pump cubicle wall connections. In addition, the analysis concluded that if the structural capacity of the steam generator and reactor coolant pump cubicle wall connections were exceeded, that this may lead to a deflection of the platform beyond anticipated design criteria, which could impact safety-related equipment within the steam generator and reactor coolant pump cubicles. Subsequent follow-up analyses determined that the platforms would maintain a safety margin of at least 1.0, which was the minimum margin for operability, but was less than the desired safety margin of 2.0. Therefore, any deflection that occurred would not affect any safety-related equipment.

To address this issue, the licensee installed support hangers to supplement the rigid supports in the event of a failure of the rigid supports. The inspectors reviewed the corrective actions that the licensee implemented to improve the overall stability of the platforms. In particular, the inspectors reviewed the engineering change, implementing work order, and held discussions with Engineering personnel to assess the licensees corrective actions. No concerns were identified.

Observation: Multiple Unit 2 Rod F-6 Failed Fuses 71152 The inspectors performed a review of plant issues, particularly those entered into the licensees corrective action program, associated with Unit 2 Control Rod F-6 failed fuses. On December 14, 2018, while performing required control rod testing, Control Rod F-6 dropped 19 steps into the core while attempting to withdraw the control rod. The licensee determined that both of the Control Rod F-6 movable gripper fuses failed due to a system ground.

Troubleshooting efforts failed to identify a cause for the system ground, which was no longer present during the troubleshooting. To address the issue, the licensee implemented a modification to install fuses rated for 15 amperes of current to replace the original fuses that had only been rated for 10 amperes of current. Subsequently, on August 12, 2019, while performing required control rod testing, Control Rod F-6 dropped fully into the core while attempting to insert the control rod. Again, the licensee determined both fuses on the movable gripper had failed due to a system ground. Troubleshooting efforts again failed to identify a cause for the system ground, which was no longer present during the troubleshooting. The licensee subsequently implemented a modification to install fuses rated for 25 amperes of current to replace the fuses rated for 15 amperes of current.

The inspectors reviewed the corrective actions that the licensee implemented to improve the reliability of the rod control system. The licensee installed fuses that were less likely to fail following a ground and installed equipment to monitor the rod control system prior to and during any planned rod motion associated with Control Rod F-6. The inspectors reviewed the temporary modifications and implementing work orders, and held follow-up discussions with Operations and Engineering personnel to assess the licensee's actions. No concerns were identified.

Minor Violation 71153 Minor Violation: On April 23, 2019, licensee personnel identified that Unit 1 and Unit 2 Reactor Coolant System (RCS) temperature had decreased below the minimum temperature limit prescribed by the Pressure and Temperature Limit Report (PTLR) on numerous occasions since 2001. These PTLR temperature limits are in place, in part, to protect the stressed region of the reactor pressure vessel (RPV) closure head materials and to protect the embrittled RPV beltline region of the reactor vessel that surrounds the core region during normal operation and transients. In 2001, the licensee adopted Improved Technical Specifications (ITS), which included Technical Specification (TS) Limiting Condition for Operation (LCO) 3.4.3, which required that the licensee maintain RCS temperature, pressure, heatup, and cooldown rates within the PTLR curves at all times and that if LCO 3.4.3 was not met in Mode 5, to: 1) immediately initiate actions to restore the parameter to within limits, and 2) to determine whether the RCS is acceptable for continued operation prior to entering Mode 4. The licensee determined Unit 1 was below the minimum temperature prescribed by the PTLR during 11 of the previous 12 refueling outages while in Mode 5 (Cold Shutdown)and that Unit 2 was below the minimum temperature prescribed by the PTLR for all of the previous 12 outages while in Mode 5. The licensee failed to perform either of the required actions for any of these occurrences. Following discovery, the licensee evaluated each of the occurrences and determined them to be acceptable utilizing the guidance of American Society of Mechanical Engineers (ASME) Boiler Pressure Vessel Code,Section XI, Appendix E, Evaluation of Unanticipated Operating Events. The licensee entered this issue into their corrective action program as AR 2309642.

Screening: The inspectors determined the performance deficiency was minor. The inspectors evaluated the condition and determined the performance deficiency and associated violation of NRC requirements was not more-than-minor after answering No to all of the questions in Block 3 of IMC 0612, Appendix B, Issue Screening.

Enforcement:

The licensee implemented actions to restore compliance. The failure to comply with Technical Specifications 3.4.3 constituted a minor violation that is not subject to enforcement action in accordance with the NRCs Enforcement Policy.

EXIT MEETINGS AND DEBRIEFS

The inspectors verified no proprietary information was retained or documented in this report.

• On September 30, 2019, the inspectors presented the integrated inspection results to Mr. R. Craven, Site Director, and other members of the licensee staff.

• On September 26, 2019, the inspectors presented the Radiation Protection inspection results to Mr. R. Craven, Site Director, and other members of the licensee staff.

DOCUMENTS REVIEWED

Inspection Type Designation Description or Title Revision or

Procedure Date

71111.01 Calculations 2002-0003 Service Water System Design Basis 7

Corrective Action AR 2313105 CWPH Damper Has Two Broken Linkages Out of Several 05/03/2019

Documents AR 2321172 W-1A CWPH Exhaust Fan is Not Running and Damper is 07/16/2019

Closed

Corrective Action AR 2321575 NRC Identified Issue with WO 40622329 (HX-269) 07/18/2019

Documents

Resulting from

Inspection

Procedures NP 7.2.29 External Events Program 8

OI 168 Emergency Diesel Generator Operability 28

OP-AA-102-1002 Seasonal Readiness 29

Work Orders 40622329 HX-269 Inspect and Clean Air Handling Unit 05/09/2019

40630828 01 PC 49 Part 6 Securing From Cold Weather Ops 06/04/2019

71111.04Q Corrective Action AR 2316067 Tap Settings For B52-DB25-026 Were Found to be Incorrect 05/29/2019

Documents

Drawings M-207 SH 1 Service Water 89

M-207 SH 1A Service Water 42

M-207 SH 3 Service Water 74

M-209 SH. 12 Emergency Diesel Generator Air Starting System 23

M-217 SH 1 Auxiliary Feedwater System 104

M-217 SH. 2 Auxiliary Feedwater System 38

M-219 SH. 1 Fuel Oil System 51

M-219 SH. 2 Fuel Oil System Diesel Generator Building 16

Procedures 0-SOP-SW-105 South Service Water Pump Header Isolation 14

CL 10D Fuel Oil Systems 24

CL 11A G-01 G-01 Diesel Generator Checklist 28

CL 13E Part 1 Auxiliary Feedwater Valve Lineup Turbine-Driven Unit 2 30

CL 13E Part 1 Auxiliary Feedwater Valve Lineup Turbine-Driven Unit 2 30

71111.05Q Corrective Action AR 2329436 U1 TH Permanent Combustible Storage Exceeding Limit 09/26/2019

Documents

Resulting from

Inspection Type Designation Description or Title Revision or

Procedure Date

Inspection

Drawings M-208 S

H. 9 Fire Protection System 1,2-X04 Transformers/Gas Turbine 3

Building

Fire Plans PFP-0-PAS Pre-Fire Plan Protected Area South 1

PFP-1-TB 26 Pre-Fire Plan Unit 1 Turbine Hall Building Elevation 26 Feet 1

PFP-1-TB 8 Pre-Fire Plan Unit 1 Turbine Hall Building Elevation 8 Feet 0

Miscellaneous Door Transaction History - Door 27 & 28 09/24/2019 -

09/26/2019

Fire Round Performance Sheet - Turbine Hall and 09/24/2019 -

Miscellaneous Areas 09/25/2019

FPTE 2016-018 PBNP Detailed Fire Modeling Report - Fire Compartment 1

301GRP Unit 1 Turbine Building General Area

FPTE 2016-024 PBNP Detailed Fire Modeling Report - Fire Compartment 1

318 Cable Spreading Room, Elevation 26'-0

Procedures PC 73 Part 2 Monthly Surveillance of Fire Hose Stations 27

PFP-0-CB Pre-Fire Plan Control Building - Elevation 8 Feet, 26 Feet, 44 1

Feet, and 66 Feet

RMP 9011-1 NFPA 805 Fire, Flood, and HELB Door Inspections 27

RMP 9057 Fire Barrier Penetration Fire Seal Surveillance 33

Work Orders 40524792 01 Inspect/Repair G-01 VNDG Dampers 07/31/2018

71111.06 Corrective Action AR 2326196 Cable Condition Monitoring Program - PMS for Visual 08/29/2019

Documents Inspect

Corrective Action AR 2329730 Gap Identified in Confined Space Entry Requirements 09/30/2019

Documents AR 2329904 NRC Questioned Sump Pump Located in Switchyard 10/01/2019

Resulting from Manhole #8

Inspection AR 2329928 Blown Fuse Found in C-324 for Controller and Sump Pump 10/01/2019

AR 2330019 Condition Reporting - Administrative Trend Identified By 10/02/2019

NRC

Drawings E-100 SH. 1 Electrical Plot 41

E-353401 Yard Area Diesel Generator Ductbank Plan 10

Procedures ER-AA-106 Cable Condition Monitoring Program 4

NP 7.7.4 Scope and Risk Significant Determination for the 25

Maintenance Rule

Inspection Type Designation Description or Title Revision or

Procedure Date

71111.11Q Corrective Action AR 2323545 Licensed Operator Simulator Evaluation Failure 08/06/2019

Documents

Miscellaneous PBN LOC 19D As Found 0

001E

Procedures 2RESP 6.1 Core Power Distribution and Nuclear Power Range Detector 6

Calibration Unit 2

OP 3A Unit 2 Power Operation to Hot Standby Unit 2 17

71111.12 Corrective Action AR 2322481 Screening of CRs For PB Systems Entering Maintenance 07/26/2019

Documents Rule (a)(1) Status

Miscellaneous Maintenance Rule (a)(3) Assessment Data Report 01/01/2017 -

06/30/2018

Procedures ER-AA-100-2002 Maintenance Rule Program Administration 7

Self-Assessments 2277705 PBN Maintenance Rule (a)(3) Assessment 2018 08/28/2018

71111.13 Corrective Action AR 2327391 Door-600 Latch Found Not Functioning 09/10/2019

Documents

Resulting from

Inspection

Miscellaneous Point Beach Unit 1 and 2 Daily Status Report 09/12/2019

Current Risk Summary Reports; PBN Unit 1 and 2 09/09/2019 -

09/10/2019

Current Risk Summary Reports; PBN Unit 1 and 2 08/12/2019 -

08/15/2019

Current Risk Summary Reports; PBN Unit 1 and 2 07/24/2019 -

07/25/2019

Point Beach Unit 1 and 2 Daily Status Report 07/24/2019

Current Risk Summary Reports; PBN Unit 1 and 2 07/11/2019 -

07/12/2019

Current Risk Summary Reports; PBN Unit 1 and 2 07/18/2019

Procedures 0-SOP-VNBI-003 White/Yellow Battery and Inverter Room Ventilation Normal 14

Operation

IT 45 Train B Safety Injection Valves Train B Unit 2 8

PC 49 Part 6 Securing From Cold Weather 32

71111.15 Corrective Action AR 2303642 Clarification Required Regarding Implementation of LCO 04/09/2019

Documents 3.4.3

Inspection Type Designation Description or Title Revision or

Procedure Date

AR 2320943 G-02: Step Change in G2H Vibration Readings 07/14/2019

AR 2328931 G-03 Sliding Feet Bolt and Hole Binding Indications 09/23/2019

AR 2329437 G-03 Exhaust Pipe Penetration HB-29-3 Inspection 09/26/2019

Operability AR 2304733 M-O-14-774-N06, G-04 Exhaust Pipe Support HB-29-8 03/07/2019

Evaluations Degraded

71111.18 Calculations 2018-0001 Unit 1 and Unit 2 Steam Generator and Reactor Coolant 0

Pump Loop Platforms

Corrective Action AR 2315473 Vendor Calculation Error 05/23/2019

Documents

Corrective Action AR 2321085 NRC Identified Calculation Error 07/15/2019

Documents

Resulting from

Inspection

Drawings C-300 Turbine Building and Primary Auxiliary Building Column 6

Location Plans

C-315 Containment Structure Interior Floor Plan at Elevation 21'-0 10

Engineering EC 292242 Contingent Supports for Unit 1 Steam Generator/Reactor 0

Changes Coolant Pump A & B Platforms

EC 292242 Contingent Supports for Unit 1 Steam Generator/Reactor 1

Coolant Pump A & B Platforms

Miscellaneous SCR 2018-0175 10 CFR 50.59 Screening: Contingent Supports for Unit 1 0

Steam Generator/Reactor Coolant Pump A & B Platforms

71111.19 Calculations 2015-02221 Flow Delivered to TDAFW Pumps Via the DDFP 0

Corrective Action AR 2322868 Bolts Loosened on P-035B-E During 30 Minute Run 07/30/2019

Documents AR 2327580 1P-2C CHG Pump Check Valves Potentially Missing 09/11/2019

Complete IST

Corrective Action AR 2327364 NRC Identified Expired Temporary Structure Permits in Yard 09/10/2019

Documents AR 2327770 Work Documentation Question on WO 40644432 (NRC 09/12/2019

Resulting from Identified)

Inspection AR 2328852 Door Issues In G-03 BLDG 09/23/2019

AR 2329888 May 2018 Work Order Missing PMT 10/01/2019

Drawings M-209 SH. 14 Starting & Service Air System Diesel Generator Building 13

Procedures OM 3.27 Control of Fire Protection and NFPA 805 Equipment 69

Work Orders 40496887 1P-002A Overhaul Pump Check Valves Per RMP 9003-9 05/22/2018

Inspection Type Designation Description or Title Revision or

Procedure Date

40583319 Inspection and Maintenance 09/23/2019 -

09/24/2019

40610426 DA-323, Disassemble, Clean, and Inspect Check Valve 09/23/2019 -

09/24/2019

40611545 2 Year Mechanical Maintenance Items 09/06/2019 -

09/13/2019

40611547 G-05, 2 Year Electrical Inspection 09/03/2019 -

09/13/2019

40626475 SW-02912-BS Inspect Zurn Strainer 07/25/2019

40626680 P-035B-E 4 Year Maintenance 07/30/2019

40629792 1P-011A Grease Coupling 08/20/2019

40629793 1P-011A Change Oil, Flush Bearings and Clean Intake 08/20/2019

40630616 IT 12 Train A 08/20/2019

40634592 PC-29; G-05 Gas Turbine Generator Load Testing 09/12/2019 -

09/13/2019

40644432 1P-002C Overhaul Pump Check Valves Per RMP 9003-9 08/27/2019 -

09/11/2019

40679325 1P-002C, Replace Pump Seals 08/29/2019

71111.22 Corrective Action AR 2307102 1RC-575A Does Not Go Open 03/24/2019

Documents AR 2309906 1RC-575A Failure of Position Indication Not Recognized as 04/10/2019

IST Failure

AR 2316834 IST Required Test Frequency Missed 2CC-779A 06/04/2019

AR 2326748 Packing Adjustment Needed During Annual Z-2204A Run 09/05/2019

Corrective Action AR 2326537 WO Title and Periodicity Not Aligned (NRC Identified) 09/03/2019

Documents AR 2327000 Procedure Use and Adherence Issue During Test 09/06/2019

Resulting from

Inspection

Drawings 10665 CD-2 Wiring Diagram - Interconnect Reactor Control System Rack 7

Sheet 2 2R1 (2C111) Bottom

617F354 Sheet 1 Reactor Protection System System Notes 14

617F354-2 Sheet Test Circuit Reactor Protection System Train B Point 2

2B Beach N.P. Unit 2

617F354-2 Sheet Reactor Protection System Train B Point Beach N.P. Unit 2 6

4B1

Inspection Type Designation Description or Title Revision or

Procedure Date

617F354-2 Sheet Reactor Protection System Train B Point Beach N.P. Unit 2 2

4B2

617F354-2 Sheet Reactor Protection System Reactor Trip Breaker Switchgear 5

5B Train B Point Beach N.P. Unit 2

617F354-2 Sheet Reactor Protection System Train B Point Beach N.P. Unit 2 1

8B2

CD1-15-1 Connection Diagram Rack 1C171B-F/1C197 2

E-2094 Sheet Panel Layout, Front View Reactor Protection, Train B Test 1

70R Cabinet 2C-165-Front Point Beach N.P. Unit 2

Engineering SCR 2014-0051 Revision to IT 07G Following SW-2891 Replacement 8

Evaluations

Miscellaneous Inservice Testing PBNP Inservice Testing Program 5th Interval 8

Program

Document

Procedures 0-SOP-FLEX-004 FLEX Portable Diesel Steam Generator Injection Pump 6

Operation Z-2004A & Z-2004B

2ICP 02.003B Reactor Protection System Logic Train B 31 Day 16

Surveillance Test

AD-AA-100-1006 Procedure and Work Instruction Use and Adherence 16

STPT 14.11 Auxiliary Feedwater Setpoint Document 29

Work Orders 40624802 01 1ICP-2.32 - 1P-29 Auxiliary Feedwater Suction Pressure 08/05/2019

Trip

40630616 IT 12 Train A 08/20/2019

40630617 IT-07G, Service Water MOV Testing 08/21/2019

71114.06 Miscellaneous EPIP 2.1 Nuclear Accident Reporting System Form (NARS) - ALERT 08/05/2019

B (Drill)

PBN LOC 19D As Found 0

001E

71124.01 Corrective Action AR 02310391 Unposted High Radiation Area Identified 04/12/2019

Documents

71124.02 Miscellaneous U1R38 Radiation Protection Post Outage Report 0

Radiation Work 19-1053 Unit 1 Narrow Range Resistance Temperature Detector 3

Permits (RWPs) Replacement

19-1056 Remove/Reinstall Reactor Vessel Head 0

Inspection Type Designation Description or Title Revision or

Procedure Date

19-1057 Steam Generator Eddy Current Testing 0

19-1064 Unit 1B Reactor Coolant Pump Impeller Change Out and 1

Pump Bowl Inspection

Self-Assessments AR 02299965-04 Level 1 NRC Baseline RP Inspection 08/19/2019

71124.03 Corrective Action AR 02326137 SCBA As Found Condition 08/28/2019

Documents AR 02326726 SCBA Found with Empty Cylinder 09/05/2019

Miscellaneous Day Shift Operators SCBA Qualifications 09/25/2019

Assignment 45- Perform Walkdown of SCBA Bottles 10/02/2019

Work Orders 40376510 SCBA Air Cylinder Hydrostatic Test 1

71151 Corrective Action CAP Search; Systems: 4160 Volt System, 480 Volt System, 07/01/2018 -

Documents Diesel Air, Diesel Generator, Fuel Oil, Diesel Generator 06/30-2019

Ventilation

Miscellaneous Reactor Oversight Program Performance Indicators (PI) and 08/01/2018 -

Monthly Operating Report Data; Various 08/01/2019

MSPI Derivation Reports; High Pressure Injection, 07/01/2018 -

Emergency AC, Cooling Water; Unavailability Index; Units 1 06/30/2019

and 2

MSPI Derivation Reports; High Pressure Injection, 07/01/2018 -

Emergency AC, Cooling Water; Unreliability Index; Units 1 06/30/2019

and 2

MSPI Margin Reports; Units 1 and 2 07/01/2018 -

06/30/2019

PBN-BFJR-18- MSPI Basis Document for PBNP 26

054

71152 Calculations 027145-CALC-06 Thermal Analysis of Unit 1 Steam Generator Cubicles 'A' 0

and 'B' Manway Platforms

PBN-BFJR-19- Point Beach Safety Significance of New Failure Modes for 0

20 Swing Battery Chargers

PBN-BFJR-19- PBN Risk Impact of Spurious Breaker Operation Due to Fire 0

033

Corrective Action AR 2156298 Platforms in All RCP and SG Cubicles Lack Documentation 09/16/2016

Documents AR 2286174 Thermal Loading from Design Basis Accident Exceeds 10/16/2018

Platform

Inspection Type Designation Description or Title Revision or

Procedure Date

AR 2294755 Control Rod F-06 Experienced Uncontrolled Motion 12/14/2018

AR 2297419 Unit 2 Manway Platforms Thermal Loading Results 01/10/2019

AR 2302088 Failure Mode Not Properly Considered for Battery Chargers 02/14/2019

AR 2319142 Failure Mode Consequences Not Properly Considered in 06/25/2019

NSCA AN

AR 2324133 Dropped F-6 Rod 08/12/2019

Drawings PBE-7033 Simplified Electrical Power Distribution Diagram PBNP Unit 13

1&2

Engineering P3257-1000-00 Independent Evaluation of AR 02302088 0

Evaluations

Miscellaneous FPPDD Fire Protection Program Design Document 2

71153 Corrective Action AR 2293462 1CW-3502 Valve Operator Broke - Unit 1 Reactor Trip 12/05/2018

Documents AR 2309642 Clarification Required Regarding Implementation of LCO 04/09/2019

3.4.3

Procedures AOP-17A U2 Rapid Power Reduction 24

AOP-6A U2 Dropped Rod 23

AOP-6H Quadrant Power Tilt 5

2