IR 05000458/2020001

Integrated Inspection Report 05000458/2020001

ML20119A851
Person / Time
Site:	River Bend
Issue date:	04/29/2020
From:	Jason Kozal NRC/RGN-IV/DRP
To:	Vercelli S Entergy Operations
References
IR 2020001
Download: ML20119A851 (21)

Text

April 29, 2020

SUBJECT:

RIVER BEND STATION - INTEGRATED INSPECTION REPORT 05000458/2020001

Dear Mr. Vercelli:

On March 31, 2020, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at River Bend Station. On April 7, 2020, 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.

Three findings of very low safety significance (Green) are documented in this report. Two of these findings involved violations of NRC requirements. We are treating these violations as non-cited violations (NCVs) consistent with Section 2.3.2 of the Enforcement Policy.

If you contest the violations or the significance or severity of the violations 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 IV; the Director, Office of Enforcement; and the NRC Resident Inspector at River Bend Station.

If you disagree with a cross-cutting aspect assignment or a finding not associated with a regulatory requirement in this report, you should provide a response within 30 days of the date of this inspection report, with the basis for your disagreement, to the 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 River Bend Station. 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 Title 10 of the Code of Federal Regulations 2.390, Public Inspections, Exemptions, Requests for Withholding.

Sincerely, Jason W. Kozal, Chief Reactor Projects Branch C Division of Reactor Projects

Docket No. 05000458 License No. NPF-47

Enclosure:

As stated

Inspection Report

Docket Number:

05000458

License Number:

NPF-47

Report Number:

05000458/2020001

Licensee:

Entergy Operations, Inc.

Facility:

River Bend Station

Location:

St. Francisville, Louisiana

Inspection Dates:

January 1, 2020 to March 31, 2020

Inspectors:

D. Antonangeli, Health Physicist

B. Baca, Health Physicist

R. Kumana, Senior Resident Inspector

B. Parks, Resident Inspector

Approved By:

Jason W. Kozal, Chief

Reactor Projects Branch C

Division of Reactor Projects

SUMMARY

The U.S. Nuclear Regulatory Commission (NRC) continued monitoring the licensees performance by conducting an integrated inspection at River Bend Station, 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.

List of Findings and Violations

Licensee Contractor Failed to Follow Radiation Work Permit Instructions for Entry into a High Radiation Area Cornerstone Significance Cross-Cutting Aspect Report Section Occupational Radiation Safety Green NCV 05000458/2020001-01 Open/Closed

[H.2] - Field Presence 71124.01 The inspectors reviewed a self-revealed, Green, non-cited violation of Technical Specifications 5.7.1 for the failure to control access to high radiation areas by means of a radiation work permit. Specifically, a licensee contractor entered a high radiation area without being on the proper radiation work permit and failed to follow the radiation work permit instructions when they received a dose rate alarm.

Failure to Follow Shutdown Safety Management Program Resulting in Both Diesel Generators Inoperable Cornerstone Significance Cross-Cutting Aspect Report Section Mitigating Systems Green FIN 05000458/2020001-02 Open/Closed

[H.5] - Work Management 71152 The inspectors identified a Green finding when the licensee failed to follow Procedure EN-OU-108,

Shutdown Safety Management Program (SSMP). Specifically, the licensee failed to ensure the outage schedule would maintain defense in depth for electrical power, which resulted in a condition where no source of onsite AC electrical power was operable for several hours while in Mode 5.

Manual Reactor Scram Initiated in Response to Loss of Feedwater due to Abnormal Plant Configuration Cornerstone Significance Cross-Cutting Aspect Report Section Initiating Events Green NCV 05000458/2020001-03 Open/Closed

[H.11] -

Challenge the Unknown 71153 The inspectors reviewed a self-revealed, Green, non-cited violation of Technical Specification 5.4.1.a for the licensees failure to maintain a procedure required by Regulatory Guide 1.33, Revision 2,

Appendix A, dated February 1978. Specifically, the licensee failed to maintain required precautions in the procedure for operation of the feedwater system. As a result, operators improperly aligned the plant after a scram and caused reactor vessel level to fall below the Level 3 setpoint, introducing a second scram.

Additional Tracking Items

Type Issue Number Title Report Section Status LER 05000458/2019-002-01 Reactor Coolant Pressure Boundary Leakage Due to Contaminates Deposited on 71153 Closed

External Pipe Surface After Legacy Adhesive Spill LER 05000458/2019-003-00 Manual Reactor Scram Initiated in Response to Loss of Feedwater due to Abnormal Plant Configuration.

71153 Closed

PLANT STATUS

River Bend Station began the inspection period at rated thermal power. On January 3, 2020, the station conducted a downpower to approximately 65 percent power to conduct a control rod sequence exchange. The unit was returned to rated thermal power on January 9, 2020. On February 10, 2020, the station reduced power to approximately 90 percent power in response to the unexpected trip of a condensate pump. The unit was returned to rated thermal power on February 14, 2020. The unit remained at 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. From January 1 - March 19, 2020, 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.

Starting on March 20, 2020, in response to the National Emergency declared by the President of the United States on the public health risks of the coronavirus (COVID-19), resident inspectors were directed to begin telework and to remotely access licensee information using available technology. During this time the resident inspectors performed periodic site visits each week and during that time conducted plant status activities as described in IMC 2515, Appendix D; and observed risk significant activities when warranted. In addition, resident and regional baseline inspections were evaluated to determine if all or a portion of the objectives and requirements stated in the IP could be performed remotely. If the inspections could be performed remotely, they were conducted per the applicable IP. In the cases where it was determined the objectives and requirements could not be performed remotely, management elected to postpone and reschedule the inspection to a later date.

71111.04 - Equipment Alignment

Partial Walkdown Sample (IP Section 03.01) (2 Samples)

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

(1) Division I standby service water system on February 20, 2020

(2) Instrument air system train B on March 25, 2020

71111.05 - Fire Protection

Fire Area Walkdown and Inspection Sample (IP Section 03.01) (2 Samples)

The inspectors evaluated the implementation of the fire protection program by conducting a walkdown and performing a review to verify program compliance, equipment functionality, material condition, and operational readiness of the following fire areas:

(1) Standby cooling tower pump A room, fire area PH-1/Z-1, on February 20, 2020

(2) Battery 1A room, fire area C-18, on March 30, 2020

71111.11Q - Licensed Operator Requalification Program and Licensed Operator Performance

Licensed Operator Requalification Training/Examinations (IP Section 03.02) (1 Sample)

(1) The inspectors observed and evaluated licensed operator requalification training on February 4, 2020.

71111.12 - Maintenance Effectiveness

Maintenance Effectiveness (IP Section 03.01) (2 Samples)

The inspectors evaluated the effectiveness of maintenance to ensure the following structures, systems, and components (SSCs) remain capable of performing their intended function:

(1) Functional failure review of reactor feedwater system on January 15, 2020

(2) Functional failure review of control building air conditioning system on February 3, 2020

71111.13 - Maintenance Risk Assessments and Emergent Work Control

Risk Assessment and Management Sample (IP Section 03.01) (4 Samples)

The inspectors evaluated the accuracy and completeness of risk assessments for the following planned and emergent work activities to ensure configuration changes and appropriate work controls were addressed:

(1) Yellow risk during surveillance testing on reactor core isolation cooling system on January 16, 2020

(2) Yellow risk during surveillance testing on Division II emergency diesel generator on January 21, 2020

(3) Yellow risk during reactor core isolation cooling maintenance and condensate pump outage on February 11, 2020

(4) Yellow risk during Division III bus outage on February 25, 2020

71111.15 - Operability Determinations and Functionality Assessments

Operability Determination or Functionality Assessment (IP Section 03.01) (4 Samples)

The inspectors evaluated the licensees justifications and actions associated with the following operability determinations and functionality assessments:

(1) Division I emergency diesel generator after associated control room vent supply fan failed to automatically start as designed during testing on January 8, 2020 (CR-RBS-2020-00058)

(2) Spent fuel pool operability with compensatory measures due to Boraflex degradation on March 24, 2020 (CR-RBS-2018-04650)

(3) Containment operability with failed motor operated valve on March 30, 2020 (CR-RBS-2020-00181)

(4) Division II main steam positive leakage control system after backseating of B21-MOVF098A to address packing leakage on March 31, 2020 (CR-RBS-2020-00094)

71111.18 - Plant Modifications

Temporary Modifications and/or Permanent Modifications (IP Section 03.01 and/or 03.02) (1 Sample)

The inspectors evaluated the following permanent modification:

(1) Control building chiller system to establish that if chilled water pumps fail to start, operations will be able to restore the control building chiller system to service before exceeding the design limit for loss of heating, ventilation, and air conditioning on March 11, 2020

71111.19 - Post-Maintenance Testing

Post-Maintenance Test Sample (IP Section 03.01) (3 Samples)

The inspectors evaluated the following post-maintenance test activities to verify system operability and functionality:

(1) Work order 52864225 following Division II emergency diesel generator outage on February 24, 2020

(2) Work order 52702801 following replacement of standby switchgear under voltage relay on March 25, 2020

(3) Work order 51295501 following inverter 1B1 capacitor replacement on March 31, 2020

71111.22 - Surveillance Testing

The inspectors evaluated the following surveillance tests:

Surveillance Tests (other) (IP Section 03.01)

(1) STP-051-4525, Revision 7, ECCS/RCIC-Reactor Vessel Water Level-Low Low, Level 2; Low Low Low, Level 1; Channel Functional Test, on January 16, 2020

(2) STP-052-6301, Revision 305, Control Rod Drive Quarterly Operability Test, on March 25, 2020

Inservice Testing (IP Section 03.01) (1 Sample)

(1) STP-256-6303, Revision 24, Standby Service Water A Loop Quarterly Pump and Valve Operability Test, on February 6,

71124.02 - Occupational ALARA Planning and Controls

Radiological Work Planning (IP Section 03.01) (5 Samples)

The inspectors evaluated the integration of as low as is reasonably achievable planning into the following work activities:

(1) Radiation Work Permit (RWP) 2019-1602, Operations Activities (Tagging, In Service Testing (IST), Local Leak Rate Testing (LLRT)) in High Radiation & Locked High Radiation Areas

(2) RWP 2019-1626, Valve Work (Air Operated Valve (AOV)/Motor Operated Valve (MOV)/Integrated Valve Team (IVT)) in High Radiation & Locked High Radiation Areas, excluding Drywell

(3) RWP 2019-1726, Valve Work (AOV/MOV) Medium and High Risk Activities in high radiation areas

(4) RWP 2019-1800, Refuel Floor Activities and RWP 2019-1801, Refuel Floor Fuel Movements, In Vessel Visual Inspection (IVVI) and Supporting Activities in high radiation areas

(5) RWP 2019-1917, Under Vessel Activities in high radiation and locked high radiation areas

Verification of Dose Estimates and Exposure Tracking Systems (IP Section 03.02) (5 Samples)

The inspectors evaluated dose estimates and exposure tracking. The inspectors reviewed the following selected and associated radiation work permits:

(1) Radiation Work Permit (RWP) 2019-1602, Operations Activities (Tagging, In Service Testing (IST), Local Leak Rate Testing (LLRT)) in High Radiation & Locked High Radiation Areas

(2) RWP 2019-1626, Valve Work (Air Operated Valve (AOV)/Motor Operated Valve (MOV)/Integrated Valve Team (IVT)) in High Radiation & Locked High Radiation Areas, excluding Drywell and RWP 2019-1736, Shielding Medium and High Risk Activities

(3) RWP 2019-1715, Emergent Medium and High Risk Activities

(4) RWP 2019-1800, Refuel Floor Activities, RWP 2019-1801, Refuel Floor Fuel Movements, In Vessel Visual Inspection (IVVI) and Supporting Activities, and RWP 2019-1917, Under Vessel Activities

(5) RWP 2019-1909, Drywell Scaffolding Activities, RWP 2019-1915, Drywell Emergent Work Activities, and RWP 2019-1932, Drywell Snubber Activities

71124.04 - Occupational Dose Assessment

Source Term Characterization (IP Section 03.01) (1 Sample)

(1) The inspectors evaluated licensee performance as it pertains to radioactive source term characterization.

External Dosimetry (IP Section 03.02) (1 Sample)

(1) The inspectors evaluated licensee performance as it pertains to external dosimetry used to assign occupational dose; to include multiple dosimeters for dose gradients.

Internal Dosimetry (IP Section 03.03) (2 Samples)

The inspectors evaluated the following internal dose assessments for actual internal exposures:

(1) Six selected individual iodine internal dose assessments due to a refueling outage airborne iodine excursion in the drywell dated March 30, 2019. This event was documented in CR-RBS-2019-01783. The maximum dose assigned was 19 millirem.

(2) An individual with facial contamination identified after removing Transversing Incore Probe (TIP) tubing under the reactor pressure vessel head dated April 2, 2019. This event was documented in CR-RBS-2019-02325. The maximum dose assigned was 23.2 millirem.

Special Dosimetric Situations (IP Section 03.04) (2 Samples)

The inspectors evaluated the following special dosimetric situations:

(1) Dose assessments for three declared pregnant workers

(2) Skin dose assessment for a discrete particle on a Westinghouse contractor performing refuel floor activities in the reactor building carousel area dated May 5, 2019. This event was documented in Personnel Contamination Event Log number 2019-24.

OTHER ACTIVITIES - BASELINE

===71151 - Performance Indicator Verification

The inspectors verified licensee performance indicators submittals listed below:

IE01: Unplanned Scrams per 7000 Critical Hours Sample (IP Section 02.01)===

(1) December 2018 through December 2019 IE03: Unplanned Power Changes per 7000 Critical Hours Sample (IP Section 02.02) (1 Sample)

(1) December 2018 through December 2019

71152 - Problem Identification and Resolution

Annual Follow-up of Selected Issues (IP Section 02.03) (2 Samples)

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

(1) Diesel generator surveillance resulting in inoperability while credited for shutdown safety function under Condition Report CR-RBS-2019-02643 on March 1, 2020

(2) Manual reactor scram initiated in response to loss of feedwater due to abnormal plant configuration under Condition Report CR-RBS-2019-03891 on March 31, 2020

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 05000458/2019-002-01, Reactor Coolant Pressure Boundary Leakage Due to Contaminates Deposited on External Pipe Surface After Legacy Adhesive Spill (ADAMS Accession No. ML19241A502)

The inspectors determined that the cause of the condition described in the LER was not reasonably within the licensees ability to foresee and correct and therefore was not reasonably preventable. No performance deficiency nor violation of NRC requirements was identified.

(2) LER 05000458/2019-003-00, Manual Reactor Scram Initiated in Response to Loss of Feedwater due to Abnormal Plant Configuration (ADAMS Accession No. ML19211A854). The circumstances surrounding this LER and an associated non-cited violation are documented in the Inspection Results section of this report.

INSPECTION RESULTS

Licensee contractor failed to follow radiation work permit instructions for entry into a high radiation area.

Cornerstone Significance Cross-Cutting Aspect Report Section Occupational Radiation Safety

Green NCV 05000458/2020001-01 Open/Closed

[H.2] - Field Presence 71124.01 The inspectors reviewed a self-revealed, Green, non-cited violation of Technical Specifications 5.7.1 for the failure to control access to high radiation areas by means of a radiation work permit.

Specifically, a licensee contractor entered a high radiation area without being on the proper radiation work permit and failed to follow the radiation work permit instructions when they received a dose rate alarm.

Description:

On September 5, 2019, a licensee contractor entered a high radiation area (HRA) to walk down smoke detectors without being on the appropriate radiation work permit (RWP) and without receiving a radiological brief. Specifically, a contractor entered a posted high radiation area in the residual heat removal C pump room and received a dose rate alarm. The individual continued their work within the radiologically controlled area (RCA) for another hour before exiting. Upon exiting the RCA, the worker was unable to log out of the area and informed radiation protection personnel about the dose rate alarm. As a result of the dose rate alarm, confirmatory survey RBS-1909-00060 was conducted and verified the dose rates to be equal to or greater than 100 millirem per hour (mR/hr) at 30 centimeters within the posted high radiation area. The survey indicated a hot spot of 500 mR/hr on contact and 200 mR/hr at 30 centimeters with a general area dose rate of 80 to 100 mR/hr.

The contractor entered the RCA on RWP 2019-1004, Task 1, which allows entry into radiation areas but not high radiation areas. The dose rate alarm under RWP 2019-1004, Task 1 was 80 mR/hr. The contractors electronic alarming dosimeters peak dose rate was 94.8 mR/hr. In addition, RWP 2019-1004, Task 1 stipulated a radiation worker self-brief was acceptable if work conditions met the criteria in procedure EN-RP-100, Radiation Worker Expectations, Revision 12. Procedure EN-RP-100, Section 5.3, Step 11, does not allow entry of workers to HRAs on a radiation worker self-brief of radiological conditions. Further, EN-RP-100, Section 5.3, Step 19, requires individuals entering an HRA must be briefed and signed on the appropriate RWP. The contractor did not obtain a radiological briefing before entering an HRA and was not on an appropriate RWP for the entry.

Further, RWP 2019-1004, Task 1 specifies workers are to stop work if an unanticipated dose rate alarm is received, exit the RCA, and notify radiation protection immediately. The contractor continued to work beyond receipt of the dose rate alarm and did not promptly notify radiation protection.

Corrective Actions: The licensee restricted access to the RCA area for the individual. The licensee provided coaching to contract/supplemental personnel as well as licensee supervisors. In addition, coaching was provided to all station personnel about the event before an entry into the RCA was made. Finally, licensee supervisors were to perform a minimum of one observation of each maintenance direct report.

Corrective Action References: Condition Report CR-RBS-2019-05861

Performance Assessment:

Performance Deficiency: The failure of a licensee contractor to follow radiation work permit instructions for entry into a high radiation area is a performance deficiency.

Screening: The inspectors determined the performance deficiency was more than minor because it was associated with the Program and 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 follow RWP instructions for entry into a high radiation area impacts the licensees ability to ensure the adequate protection of the worker health and safety.

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

(1) was not related to as low as is reasonably achievable (ALARA) planning or work controls,

(2) it was not an overexposure,

(3) there was not a substantial potential for overexposure, and

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

Cross-Cutting Aspect: H.2 - Field Presence: Leaders are commonly seen in the work areas of the plant observing, coaching, and reinforcing standards and expectations. Deviations from standards and expectations are corrected promptly. Senior managers ensure supervisory and management oversight of work activities, including contractors and supplemental personnel. Specifically, the supervisor failed to perform supplemental employee oversight and engagement to reinforce expectations according to Procedure EN-MA-101, Conduct of Maintenance, Revision 31, and failed to perform required performance observations according to Procedure EN-OM-126, Contract Management, Revision 8.

Enforcement:

Violation: Technical Specification 5.7.1 requires access to high radiation areas, as defined in 10 Code of Federal Regulations (CFR) Part 20, be controlled by requiring issuance of a radiation work permit (RWP). RWP 2019-1004, Task 1 did not allow entry into high radiation areas and specified workers are to stop work if an unanticipated dose rate alarm is received, exit the RCA, and notify radiation protection immediately.

Contrary to the above, on September 5, 2019, a failure to control access to a high radiation area with the issuance of a RWP occurred. Specifically, a licensee contractor entered a high radiation area under RWP 2019-1004, Task 1, which did not allow entry into the area. Further, the contractor did not stop work when an unanticipated dose rate alarm was received and exit the RCA, nor did the contractor notify radiation protection immediately.

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

Failure to Follow Shutdown Safety Management Program Resulting in Both Diesel Generators Inoperable Cornerstone Significance Cross-Cutting Aspect Report Section Mitigating Systems

Green FIN 05000458/2020001-02 Open/Closed

[H.5] - Work Management 71152 The inspectors identified a Green finding when the licensee failed to follow Procedure EN-OU-108, Shutdown Safety Management Program (SSMP). Specifically, the licensee failed to ensure the outage schedule would maintain defense in depth for electrical power, which resulted in a condition where no source of onsite AC electrical power was operable for several hours while in Mode 5.

Description:

During the outage planning process for refueling outage RF-20, the licensee reviewed their planned schedule in accordance with Procedure EN-OU-108, Shutdown Safety Management Program (SSMP). The licensees Outage Risk Assessment Team (ORAT) reviewed the Rev 0 schedule per Section 5.4 to assess the availability of each Key Safety Function for defense-in-depth.

Procedure EN-OU-108, Attachment 9.4, Electrical Power Availability Defense in Depth, includes as part of its criteria: Ensure offsite supplies of power sources have diesel backup to ensure that sufficient power is available for needed systems.

The licensee divided the total outage schedule into four segments. The first segment covered the initial shutdown through entry into Mode 5. The second segment included the period in which the reactor cavity was flooded for core offload through the planned switch in the out-of-service equipment from one division to the other. That segment was planned for March 31 to April 25, 2019, with core offload expected on April 19, 2019. The division that was scheduled for work during that segment was Division II and the Division II diesel generator was planned to be unavailable from March 31 to April 19, 2019. The Rev 0 schedule did not contain any activities that would affect defense-in-depth of electrical power during this segment, and the Rev 0 risk assessment assumed that the Division I diesel generator would be available during the entire segment. This draft assessment was completed in February of 2019.

At some point prior to February 28, 2019, the licensee added the Division I diesel generator monthly surveillance test to the outage schedule. The licensee had an informal practice to schedule these surveillance tests immediately prior to the outage to prevent scheduling conflicts during the outage, but in this case the planners had conflicts with other planned work prior to the outage. Because of this other work, the Division I diesel generator surveillance test was last performed on March 4, 2019, and the next test was added to the outage schedule for April 8, 2019. The Division I diesel generator surveillance test is performed in accordance with licensee Procedure STP-309-0201, Division I Diesel Generator Operability Test. In accordance with this procedure, actions are taken during the test that result in periods of time when the diesel generator cannot automatically start and supply loads on the associated safety bus.

Procedure EN-OU-108, Section 5.4, Step 9, states that:

After ORAT review of the schedule (normally after issuance of Rev 0), when a schedule change involving a change to system window logic or an emergent activity is added, the outage organization is responsible for the following:

Perform a Higher Risk Evolution Review, if applicable.

Review the change or emergent activity for impact on shutdown safety.

Obtain approval and make notifications as applicable using Attachment 9.11, Outage Change/Emergent Activity Evaluation. Proper approvals are required prior to executing the schedule change.

However, following the addition of the Division I diesel generator surveillance test to the schedule, the ORAT was not aware of the change, and did not update the risk assessment to include this test. On March 19, 2019, the Rev 1 risk assessment was issued. This was the final assessment prior to the outage and included the assumption that the Division I diesel generator would remain available during this outage time segment. As a result, no review of the impact on shutdown safety was performed.

On April 8, 2019, the operating crew realized that the Division I test was scheduled for that morning, but the Division II diesel generator was still unavailable due to the ongoing work window and could not be immediately restored. The reactor still had fuel bundles left in the core and could not be immediately offloaded. The Division I diesel generator would exceed its maximum allowed time between surveillance tests on April 11, 2019, and would then become inoperable. The licensee determined that neither restoration of the Division II diesel generator nor completion of core offload and exit from Mode 5 could be accomplished prior to the inoperability of the Division I diesel generator due to the expired surveillance test.

The licensee postponed the test to April 11, 2019, and modified their procedure to minimize the time that the diesel would be incapable of automatically starting and powering the safety bus. They assessed the risk to core damage from a complete loss of AC power and determined that with the number of fuel bundles remaining in the core, the Division I diesel generator could be manually restored in the event of a loss of offsite power with sufficient time to restore decay heat removal prior to core damage. Based on that assessment, the licensee considered the Division I diesel generator to be available during the surveillance test. During the surveillance test, the licensee entered a condition where no diesel generator was operable while in Mode 5. In this condition, technical specifications required action to be initiated immediately to restore one diesel generator to an operable status. The licensee completed the test and restored the Division I diesel generator to an operable status within approximately two and a half hours.

Corrective Actions: The licensee restored one diesel generator to an operable status.

Corrective Action References: CR-RBS-2020-00964

Performance Assessment:

Performance Deficiency: This failure to follow licensee Procedure EN-OU-108, Shutdown Safety Management Program (SSMP), when managing risk due to maintenance activities during a refueling outage is a performance deficiency.

Screening: The inspectors determined the performance deficiency was more than minor because it was associated with the Configuration Control attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the performance deficiency resulted in both onsite power sources being inoperable while in a mode in which one was required.

Significance: The inspectors assessed the significance of the finding using Inspection Manual Chapter 0609, Appendix G, Shutdown Safety SDP. The inspectors determined that the finding occurred during plant operational state 3. The inspectors determined that the finding was of very low safety significance (Green) because the finding did not involve an actual loss of safety function of either train.

Cross-Cutting Aspect: H.5 - Work Management: The organization implements a process of planning, controlling, and executing work activities such that nuclear safety is the overriding priority. The work process includes the identification and management of risk commensurate to the work and the need for coordination with different groups or job activities. The licensee failed to ensure that the work process included the identification and management of risk commensurate to the work and the need for coordination with different groups.

Enforcement:

Inspectors did not identify a violation of regulatory requirements associated with this finding.

Reactor Scram Initiated in Response to Loss of Feedwater due to Abnormal Plant Configuration Cornerstone Significance Cross-Cutting Aspect Report Section Initiating Events

Green NCV 05000458/2020001-03 Open/Closed

[H.11] -

Challenge the Unknown 71153 The inspectors reviewed a self-revealed, Green, non-cited violation of Technical Specification 5.4.1.a for the licensees failure to maintain a procedure required by Regulatory Guide 1.33, Revision 2, Appendix A, dated February 1978. Specifically, the licensee failed to maintain required precautions in the procedure for operation of the feedwater system. As a result, operators improperly aligned the plant after a scram and caused reactor vessel level to fall below the Level 3 setpoint, introducing a second scram.

Description:

On May 31, 2019, River Bend Station commenced a planned shutdown of the reactor to resolve a leak in the A feedwater heater string. At the time, the A feedwater heater string was isolated, and the B feedwater heater string was in service.

As steam demand was reduced during the shutdown, the minimum flow valve for the operating feed pump began to open, recirculating feedwater flow back to the condenser. The opening of the minimum flow valve increased the amount of flow through the B feedwater heater string, causing the condensation on the shell side of the feedwater heaters to increase. As a result, a high-level condition occurred in the B 5th point feedwater heater, causing an automatic isolation of the B heater string, which was the only available heater string. Without a heater string flow path available, the running feedwater pumps tripped on low suction pressure, which prompted operators to manually scram the reactor due to loss of feed flow to the reactor.

Operators placed the reactor core isolation cooling (RCIC) system in service immediately after the scram occurred, which provided make up water flow to the reactor pressure vessel (RPV). RPV level was restored to normal and the emergency operating procedure for scram was exited.

Approximately 4 minutes after the scram, operators started a feedwater pump to establish the normal feed flow path. Approximately 8 minutes after the scram, operators secured the RCIC system, as it was determined by the operators to be no longer needed to maintain reactor vessel water level.

Approximately 29 minutes after the scram, the station secured the B condensate pump, leaving only one condensate pump running. Approximately 45 minutes after the scram, the station experienced an unexpected trip of the running feedwater pump on low suction pressure. Subsequent attempts to restart a feedwater pump and restore feedwater flow were unsuccessful, and RPV level fell below the Level 3 setpoint, causing a second scram to occur due to RPV low water level.

The primary cause of the feedwater pump trip was an improper alignment of the feedwater system. Both heater strings in the system were isolated, and the only suction flow path was through a line that bypasses the heater strings. The single running condensate pump was unable to provide sufficient flow through the bypass line to allow for reliable feedwater pump operation.

The station procedure for reactor feedwater system operation, SOP-0009, Reactor Feedwater System (Sys #107), Revision 80, and the station procedure that contains instructions for feedwater pump operation in emergency and transient conditions, OSP-0053, Emergency and Transient Response Support Procedure, Revision 27, did not contain any precautions to ensure the availability of a heater string prior to and during operation. Consequently, operators incorrectly believed that a heater string was not required to be in service for reliable feedwater pump operation. Operators therefore placed a feedwater pump in service as the sole means of reactor water level control without aligning a heater string, eventually causing a low suction pressure trip of the feedwater pump and a Level 3 scram actuation.

Regulatory Guide 1.33, Revision 2, Appendix A, Section 5, states that the guidelines of ANSI N18.7-1976/ANS-3.2, Section 5.3.2, which cover the content of procedures, are subsumed as requirements of Regulatory Guide 1.33. ANSI N18.7-1976/ANS-3.2, Section 5.3.2, states that each procedure shall include precautions to alert the individual performing the task to those important measures which should be used to protect equipment and personnel, including the public, or to avoid an abnormal or emergency situation. Procedures SOP-0009 and OSP-0053 did not include precautions to ensure the existence of an appropriate system flow path alignment to support feedwater pump operation. This deficiency led to an abnormal situation in which all feedwater flow was lost, and water level in the vessel fell below the Level 3 setpoint.

Corrective Actions: The licensee restored the feedwater system to a normal operating lineup and restored reactor water level to the appropriate band.

Corrective Action References: CR-RBS-2019-03891

Performance Assessment:

Performance Deficiency: The failure to maintain a procedure for operation of the feedwater system in accordance with the requirements of plant technical specifications was a performance deficiency.

Screening: The inspectors determined the performance deficiency was more than minor because it was associated with the Configuration Control attribute of the Initiating Events Cornerstone and adversely affected the cornerstone objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Specifically, the performance deficiency resulted in a loss of reactor water level control and an associated scram actuation.

Significance: The inspectors assessed the significance of the finding using Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power.

The inspectors determined that the finding was of very low safety significance (Green) because the loss of injection did not cause a reactor trip and the loss of mitigation equipment relied upon to transition the plant from the onset of the trip to a stable shutdown condition.

Cross-Cutting Aspect: H.11 - Challenge the Unknown: Individuals stop when faced with uncertain conditions. Risks are evaluated and managed before proceeding. Specifically, operators failed to question the abnormal alignment of the feedwater system when transferring reactor water level control from the RCIC system to the running feedwater pump.

Enforcement:

Violation: Technical Specification 5.4.1.a requires, in part, that written procedures shall be established, implemented, and maintained covering the applicable procedures recommended in Regulatory Guide 1.33, Revision 2, Appendix A, dated February 1978. Regulatory Guide 1.33, Revision 2, Appendix A, Section 4.o identifies procedures covering the operation of the feedwater system as recommended procedures. Contrary to the above, prior to June 1, 2019, the licensee failed to maintain written procedures covering the operation of the feedwater system. Specifically, station procedures SOP-0009 and OSP-0053 that covered feedwater system operation were not maintained to include necessary precautions to prevent operation of the system in an improper lineup.

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

EXIT MEETINGS AND DEBRIEFS

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

On February 14, 2020, the inspectors presented the exit meeting for Occupational Radiation Safety 71124.02 and 71124.04 inspection results to K. Scott, General Manager Plant Operations and other members of the licensee staff.

On April 7, 2020, the inspectors presented the integrated inspection results to S. Vercelli, Site Vice President, and other members of the licensee staff.

DOCUMENTS REVIEWED

Inspection

Procedure

Type

Designation

Description or Title

Revision or

Date

71111.15

Calculations

NEAD-SR-19/002

Evaluation of RBS SFP Rack Criticality with Installed Inserts

Corrective Action

Documents

CR-RBS-

2018-04650, 2018-04756

71124.01

Corrective Action

Documents

CR-RBS-

2019-05861, 2019-3104, 2019-3118

71124.02

Corrective Action

Documents

CR-RBS-

2019-2427, 2019-4375, 2019-4463, 2019-4714

2019-6560, 2020-0393, 2020-0590

Corrective Action

Documents

Resulting from

Inspection

CR-RBS-

20-00756, 2020-00762

Miscellaneous

RBS-RPT-19-003

Review of Radiological Data Relevant to Refuel (RF) 20

03/28//2019

Procedures

EN-FAP-RP-005

Radiation Protection Five-Year Exposure Reduction Planning

EN-RP-105

Radiological Work Permits

EN-RP-110

ALARA Program

EN-RP-110-03

Collective Radiation Exposure (CRE) Reduction Guidelines

EN-RP-110-04

Radiation Protection Risk Assessment Process

EN-RP-110-06

Outage Dose Estimating and Tracking

EN-RP-112

Alpha Monitoring

EN-RP-131

Air Sampling

Radiation

Surveys

RBS-2002-00067

Post RWCU Liner Transfer

2/06/2020

RBS-2002-00095

Pre-job Survey for RWCU resin sample collection

2/10/2020

Radiation Work

Permits (RWPs)

20191436

Shielding Activities, excluding Drywell

and 1

20191602

Operations Activities (Tagging, InService Testing, Local Leak

Rate Testing) in High Radiation & Locked High Radiation

Areas

20191626

Valve Work (Air Operated Valve/Motor Operated

Valve/Integrated Valve Team) in High Radiation & Locked

High Radiation Areas excluding Drywell

0, 1, and 2

20191636

Shielding Installation/Removal in High Radiation and Locked

High Radiation Areas, Excluding Drywell

20191715

Emergent Medium and High Risk Activities

0, 1, 2

Inspection

Procedure

Type

Designation

Description or Title

Revision or

Date

20191726

Valve Work (Air Operated Valve/Motor Operated Valve)

Medium and High Risk Activities

20191736

Shielding in Medium and High Risk Activities

20191800

Refuel Floor Activities

and 1

20191801

Refuel Floor Fuel Movements, In Vessel Visual Inspection,

and Supporting Activities

and 1

20191904

Drywell Maintenance Activities

and 1

20191909

Drywell Scaffolding Activities

and 1

20191915

Drywell Emergent Work Activities

20191917

Under Vessel Activities

and 1

20191932

Drywell Snubber Activities

20191936

Drywell Shielding Installation/Removal

20191953

Bio-Shield Activities

Self-Assessments LO-RLO-2017-

00076

Pre-NRC Occupational ALARA Planning & Controls

(71124.02)

01/10/2019

QA-14/15-2019-

RBS-1

Combined Radiation Protection and Radwaste

10/28/2019

71124.04

Corrective Action

Documents

CR-RBS-

2018-0597, 2018-3352, 2018-3535, 2018-4091,

2019-0659, 2019-0949, 2019-1439, 2019-1783,

2019-1786, 2019-2325, 2019-4747, 2019-5785,

2019-5984

Corrective Action

Documents

Resulting from

Inspection

CR-RBS-

20-00705, 2020-00744

Miscellaneous

Internal Dose Assessment due to facial contamination (CR-

RBS-2019-02325)

07/01/2019

100518-0

NVLAP Certificate of Accreditation to ISO/IEC 17025:2005

01/01/2020

EN-RPT-19-003

Standardization of a Bias Between DLRs and SRD in the

Entergy Fleet - Update Based on 2018 Results

10/31/2019

PCE 2019-24

Skin Dose Assessment for discrete particle

05/05/2019

RBS-RPT-19-004

Review of the Capabilities of the Passive Monitoring Program

to Detect Intakes at RBS

04/10/2019

RPG-M-19-008

Dose Calculations (CR-RBS-2019-1783)

04/30/2019

Inspection

Procedure

Type

Designation

Description or Title

Revision or

Date

RPG-M-19-015

Personnel DLR DDE/SRD Discrepancies - Dosimetry

Investigation Reports (CR-2019-05984)

2/30/2019

Self-Assessments LO-RLO-2017-

00055

(Pre-NRC FSA) Occupational Dose Assessment (71124.04)

10/30/2017

71152

Corrective Action

Documents

CR-RBS-

2019-02643

Miscellaneous

RF-20 Outage Report

03/18/2019

RBS RF20 Level III

01/03/2019

RBS RF20 Level III

2/28/2019

Procedures

EN-OU-108

Shutdown Safety Management Program

OSP-0037

Shutdown Operations Protection Plan (SOOP)

71153

Corrective Action

Documents

CR-RBS-

2019-03891

Procedures

OSP-0053

Emergency and Transient Response Support Procedure

SOP-0009

Reactor Feedwater System (Sys #107)

80