IR 05000313/2017007

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NRC Design Bases Assurance Inspection (Programs) Report 05000313/2017007 and 05000368/2017007
ML17265A274
Person / Time
Site: Arkansas Nuclear  Entergy icon.png
Issue date: 09/21/2017
From: Thomas Farnholtz
Region 4 Engineering Branch 1
To: Richard Anderson
Entergy Operations
References
IR 2017007
Download: ML17265A274 (19)
v • d • e


Text

ptember 21, 2017

SUBJECT:

ARKANSAS NUCLEAR ONE, UNITS 1 AND 2 - NRC DESIGN BASES ASSURANCE INSPECTION (PROGRAMS) REPORT 05000313/2017007 AND 05000368/2017007

Dear Mr. Anderson:

On August 10, 2017, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at your Arkansas Nuclear One, Units 1 and 2. The NRC inspectors discussed the results of this inspection with Mr. T. Evans, Acting Vice President, and other members of your staff. The results of this inspection are documented in the enclosed report.

NRC inspectors documented one finding of very low safety significance (Green) in this report.

This finding involved a violation of NRC requirements.

If you contest the violation or significance of the NCV, you should provide a response within 30 days of the date of this inspection report, with the basis for your denial, to the U.S. Nuclear Regulatory Commission, ATTN: Document Control Desk, Washington, DC 20555-0001; with copies to the Regional Administrator, Region IV; the Director, Office of Enforcement; and the NRC resident inspector at the Arkansas Nuclear One.

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 Arkansas Nuclear One. 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/

Thomas R. Farnholtz, Chief Engineering Branch 1 Division of Reactor Safety Docket Nos. 50-313 and 50-368 License Nos. DPR-51 and NPF-6

Enclosure:

Inspection Report 05000313/2017007 and 05000368/2017007 w/Attachment:

Supplemental Information

REGION IV==

Docket: 05000313 and 05000368 License: DPR-51 and NPF-6 Report: 05000313/2017007 and 05000368/2017007 Licensee: Entergy Operations, Inc.

Facility: Arkansas Nuclear One, Units 1 and 2 Location: Junction of Highway 64 West and Highway 333 South Russellville, Arkansas Dates: July 24 through August 10, 2017 Team Leader: J. Drake, Senior Reactor Inspector, Engineering Branch 2 Inspectors: S. Alferink, Reactor Inspector J. Braisted, Reactor Inspector Approved By: Thomas R. Farnholtz, Branch Chief Engineering Branch 1 Division of Reactor Safety Enclosure

SUMMARY

IR 05000313/2017007; 05000368/2017007 07/24/2017 - 08/10/2017; Arkansas Nuclear One,

Units 1 and 2; Inspection Procedure 71111.21N, Design Bases Assurance (Programs)

The inspection activities described in this report were performed between July 24, 2017, and August 10, 2017, by three inspectors from the NRCs Region IV office. One finding of very low safety significance (Green) is documented in this report. This finding involved a violation of NRC requirements. The significance of inspection findings is indicated by their color (Green,

White, Yellow, or Red), which is determined using Inspection Manual Chapter 0609,

Significance Determination Process dated April 29, 2015. Cross-cutting aspects are determined using Inspection Manual Chapter 0310, Aspects Within the Cross-Cutting Areas dated December 4, 2014. Violations of NRC requirements are dispositioned in accordance with the NRCs Enforcement Policy dated November 1, 2016. The NRC's program for overseeing the safe operation of commercial nuclear power reactors is described in NUREG-1649, Reactor Oversight Process.

NRC Identified and Self-Revealing Findings

Cornerstone: Mitigating Systems

Green.

The inspectors identified a Green, non-cited violation of 10 CFR Part 50,

Appendix B, Criterion XVI, Corrective Actions, requires, in part, that measures shall be established to assure that conditions adverse to quality, such as failures, malfunctions, deficiencies, deviations, defective material and equipment, and nonconformance are promptly identified and corrected. Specifically, from 1996 until August 10, 2017, the licensee failed to properly resolve the environmental conditions in room 38 following a high-energy line break, even when challenged during a self-assessment by members of the quality assurance group in June 29, 2015. In response to this issue, the licensee determined that in the event of a break in the letdown line, an engineered safety feature signal automatic closure of both the inside and outside reactor building isolation valves occurs in approximately 40 seconds, preventing room 38 from going harsh. This finding was entered into the licensees corrective action program as Condition Report CR-ANO-1-2017-02441.

The inspectors determined that the licensees failure to adequately evaluate and take prompt corrective actions to resolve an identified condition adverse to quality related to the high energy line break analysis for room 38 was a performance deficiency. The performance deficiency was more than minor because it was associated with the design control attribute of the Mitigating Systems Cornerstone and affected the associated objective to ensure availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, when the licensee identified that the environmental conditions in room 38 of the auxiliary building were harsh, as determined by Design Bases Calculation CALC-01-EQ-1002-02, they failed to properly resolve the condition adverse to quality. In accordance with Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, dated June 19, 2012, Exhibit 2, Mitigating Systems Screening Questions, dated July 1, 2012, the inspectors determined that the finding was of very low safety significance (Green) because the finding (1) was not a deficiency affecting the design and qualification of a mitigating structure, system, or component, and did not result in a loss of operability or functionality; (2) did not represent a loss of system and/or function; (3) did not represent an actual loss of function of at least a single train for longer than its allowed outage time, or two separate safety systems out-of-service for longer than their technical specification allowed outage time; and (4) does not represent an actual loss of function of one or more non-technical specification trains of equipment designated as high safety-significant for greater than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> in accordance with the licensees maintenance rule program. The finding has a cross-cutting aspect in the area of human resources, training, because the organization failed to provide training and ensure knowledge transfer to maintain a knowledgeable, technically competent workforce and instill nuclear safety values. Specifically, there was a lack of understanding of the current licensing bases for the plant displayed by engineering, operations, and management [H.9]. (Section 1R21.2.2)

Licensee Identified Violations

None

REPORT DETAILS

REACTOR SAFETY

1R21 Design Basis Assurance Inspection (Programs)

Main article: IP 71111.21N

a. Inspection Scope

The inspectors performed an inspection as outlined in NRC Inspection Procedure 71111.21N, Attachment 1, Environmental Qualification under 10 CFR 50.49, Programs, Processes, and Procedures. The inspectors assessed the implementation of the environmental qualification program as required by 10 CFR 50.49, Environmental qualification of electric equipment important to safety for nuclear power plants, by Arkansas Nuclear One, Units 1 and 2. The inspectors evaluated whether Arkansas Nuclear One, Units 1 and 2, staff properly maintained the environmental qualification of electrical equipment important to safety throughout plant life, established and maintained required environmental qualification documentation records, and implemented an effective corrective action program to identify and correct environmental qualification-related deficiencies.

The inspection included a review of environmental qualification program procedures, component environmental qualification files, environmental qualification test records, equipment maintenance and operating history, maintenance and operating procedures, vendor documents, design documents, and calculations. The inspectors interviewed program owners, engineers, maintenance staff, and warehouse staff. The inspectors performed in-plant walkdowns (where accessible) to verify equipment was installed as described in the environmental qualification component documentation files for Arkansas Nuclear One, Units 1 and 2, and that the components were installed in their tested configuration. Additionally, the inspectors performed in-plant walkdowns to determine whether equipment surrounding the components could fail in a manner that could prevent the safety functions of the components and to verify that components located in areas susceptible to a high-energy line break were properly evaluated for operation in a harsh environment. The inspectors reviewed and inspected the storage of replacement parts and associated procurement records to verify environmental qualification parts approved for installation in the plant were properly identified and controlled, and that storage and environmental conditions did not adversely affect the components qualified lives. Documents reviewed for this inspection are listed in the attachment.

In accordance with the inspection procedure, the inspectors initially selected 10 components to assess the adequacy of the environmental qualification program.

Component samples selected for this inspection were:

  • CV-1407, Unit 1, Borated Water Storage Tank T-3 Outlet Valve Operator
  • SV-3840, Unit 1, CV-3840 Control Air Isolation Solenoid Valve for Decay Heat Pump P-34A Cooling Jacket Service Water Inlet Valve
  • ZS-2613, Unit 1, Internal Switch for CV-2613 Valve Operator for Steam to Emergency Feedwater Pump Turbine P-7A
  • 2VSF-1C, Unit 2, Containment Cooler Fan Motor
  • 2VSF-1A, Unit 2, Containment Cooler Fan Motor During the course of the inspection, the following item was included in the scope:
  • Electric governor for P-7A, Unit 1, Emergency Feedwater Pump Turbine

b. Findings

Failure to Promptly Identify and Correct an Inadequate Design Bases Calculation

Introduction.

The inspectors identified a Green, non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, associated with the licensees failure to adequately evaluate and take prompt corrective actions to address an identified condition adverse to quality related to the reactor coolant system letdown line break analysis for room 38 of the auxiliary building.

Description.

While reviewing Calculation NES-13, Environmental Qualification (EQ) -

Environmental Service Conditions, Revision 13, the inspectors noted that the environment in room 38 (emergency feedwater pump room) following a letdown line break was considered harsh, with the temperature in the room exceeding 150 degrees Fahrenheit (°F) as determined by Design Bases Calculation CALC-01-EQ-1002-01, Reactor Coolant Letdown Line HELB Analysis, Revision 2. Title 10 CFR 50.49, Environmental Qualification of Electric Equipment Important to Safety for Nuclear Power Plants, requires the licensee to place safety-related electric equipment that is relied upon to remain functional during and following design basis events in the environmental qualification program. Neither of the emergency feedwater pumps or any of their subcomponents were within the scope of the environmental qualification program, even though both of them were safety-related components.

Reviewing the licensees environmental qualification program, the inspectors noted that both trains of the emergency feedwater system had temperature-sensitive components. Revision 2 of CALC-01-EQ-1002-01 indicated that the design basis letdown line break would result in the temperature of the room housing the emergency feedwater pumps (room 38) exceeding 150 °F, making it a harsh environment. Since a letdown line break could result in actuation of both emergency feedwater pumps, the ability of both the motor-driven and the turbine-driven pumps to perform their intended function could be challenged during this event.

The inspectors informed the licensee of their concern on August 7, 2017.

During follow-up discussions, the inspectors determined that the licensee had previously entered this issue into their corrective action program as Condition Report CR-ANO-1-2015-02630 in March 2015. Subsequent review of this condition report revealed that the condition report was closed with the resolution that, EFW (emergency feedwater) is not essential in this case since cooldown may be accomplished by continuance of normal main feedwater flow from either MFW pump in conjunction with the condensate system.

The inspectors reviewed the following information:

  • Calculation ER-93-R-1040-01 indicates that the emergency feedwater motor-driven pump motor bearings are qualified to a room ambient temperature of 148 °F.
  • Calculation ER-93-R-1040-01 indicates that the Woodward governor on the emergency feedwater turbine driven pump is qualified to a room ambient temperature of 150 °F.
  • Design Basis Calculation CALC-01-EQ-1002-01 states that for the design bases event of a reactor coolant system letdown line break, the temperature in room 38 exceeds 150 °F and both trains of emergency feedwater would be lost.
  • The assumptions in Calculation ER-93-R-1040-01 state that the design break in the reactor coolant letdown line will result in a net loss from the reactor coolant system of 48 pounds mass per second in excess of the reactor coolant system make up system capacity and result in a drop in pressurizer level of approximately 24 inches per minute until the leak is isolated.

Additionally, the emergency feedwater motor-driven pump motor bearings and the Woodward governor on the emergency feedwater turbine-driven pump were previously in the environmental qualification program. They were removed from the program in 1986 based on modifications to the upper south penetration room (room 77) that resulted in establishment of a new high-energy line break vent path for the main feedwater line break. As a result of the modifications, a main feedwater line break would exhaust into the boiler room from room 77 and would no longer affect room 38. At that time, the licensee determined that room 38 would be considered a mild environment.

In 1996, the licensee reevaluated the reactor coolant letdown line high energy line break and determined that room 38 would be a harsh environment based on the assumptions used in Design Bases Calculation CALC-01-EQ-1002-01, Revision 2. When this was identified, rather than place the emergency feedwater motor-driven pump motor bearings and the Woodward governor on the emergency feedwater turbine-driven pump back in the environment qualification program, the licensee made the determination that emergency feedwater was not required for this design bases event and they could rely on main feedwater and condensate to provide required cooling. This is the same justification that was used to close Condition Report CR-ANO-1-2015-02630.

The licensee did not initiate a condition report on the NRCs concern, even after the inspectors questioned if that was in accordance with their corrective action procedures.

The licensee stated that the issue did not represent an immediate safety concern because they had previously performed operability assessments for the affected areas, which established a reasonable expectation for operability pending resolution of the identified issue. After the inspectors presented the licensee with the information in the safety analysis report regarding the turbine generator and main feedwater pumps, the licensee reevaluated the assumptions used in CALC-01-EQ-1002-01, Revision 2, and determined that the assumed time to isolate the leak was overly conservative and a more appropriate time to isolate the leak was five minutes. The licensee determined that in the event of a break in the letdown line, an engineered safeguards signal automatic closure of both the inside and outside reactor building isolation valves occurs in approximately 30 to 40 seconds. The Channel 2 engineered safety feature actuation signal (engineered safety feature actuation signal) (low RCS pressure or high reactor building pressure) will also initiate closure of CV-1221 (Aux Building), and engineered safety feature actuation signal Channel 1 (low RCS pressure or high reactor building pressure) will initiate closure of CV-1214 and CV-1216 (Letdown coolers outlet Reactor building side) isolating the break flow and preventing room 38 from going harsh. The licensee determined that securing the leak within five minutes would prevent the environment in room 38 from becoming harsh for the reactor coolant letdown line break. The licensee entered the issue into the corrective action program on August 9, 2017, as Condition Report CR-ANO-1-2017-02441. Because the assumptions in the design bases calculation were determined by the licensee to be overly conservative, the equipment would not have been subjected to a harsh environment, function would not have been lost.

The inspectors determined that the licensee had failed to promptly correct a condition adverse to quality. The original acceptance by engineering, operations, and management that it was acceptable to rely on main feedwater to provide required cooling during design bases events points to a training deficiency associated with understanding the current licensing bases throughout the organization.

Analysis.

The inspectors determined that the licensees failure to adequately evaluate and take prompt corrective actions to resolve an identified condition adverse to quality related to the high energy line break analysis for room 38 was a performance deficiency.

The performance deficiency was more than minor because it was associated with the design control attribute of the Mitigating Systems Cornerstone and affected the associated objective to ensure availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, when the licensee identified that the environmental conditions in room 38 of the auxiliary building were harsh, as determined by Design Bases Calculation CALC-01-EQ-1002-02, they failed to properly resolve the condition adverse to quality. In accordance with Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, dated June 19, 2012, Exhibit 2, Mitigating Systems Screening Questions, the inspectors determined that the finding was of very low safety significance (Green) because the finding

(1) was not a deficiency affecting the design and qualification of a mitigating structure, system, or component, and did not result in a loss of operability or functionality;
(2) did not represent a loss of system and/or function;
(3) did not represent an actual loss of function of at least a single train for longer than its allowed outage time, or two separate safety systems out-of-service for longer than their technical specification allowed outage time; and
(4) does not represent an actual loss of function of one or more non-technical specification trains of equipment designated as high safety-significant for greater than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> in accordance with the licensees maintenance rule program. The finding has a cross-cutting aspect in the area of human resources, training, because the organization failed to provide training and ensure knowledge transfer to maintain a knowledgeable, technically competent workforce and instill nuclear safety values. Specifically, there was a lack of understanding of the current licensing bases for the plant displayed by engineering, operations, and management [H.9].
Enforcement.

The inspectors identified a Green, non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Actions, requires, in part, that measures shall be established to assure that conditions adverse to quality, such as failures, malfunctions, deficiencies, deviations, defective material and equipment, and nonconformance are promptly identified and corrected. Contrary to the above, from 1996, until August 10, 2017, the licensee failed to establish measures to assure that conditions adverse to quality, such as failures, malfunctions, deficiencies, deviations, defective material and equipment, and nonconformance were promptly identified and corrected. Specifically, the licensee failed to properly resolve the environmental conditions in room 38 following a high-energy line break, even when challenged during a self-assessment by members of the quality assurance group in June 29, 2015. In response to this issue, the licensee determined that in the event of a break in the letdown line, an engineered safety feature signal automatic closure of both the inside and outside reactor building isolation valves occurs in approximately 40 seconds, preventing room 38 from going harsh. This finding was entered into the licensees corrective action program as condition report CR-ANO-1-2017-02441. Because this finding was of very low safety significance and has been entered into the licensees corrective action program, this violation is being treated as a non-cited violation, consistent with Section 2.3.2.a of the NRC Enforcement Policy:

NCV 05000313/2017007-01, Failure to Promptly Identify and Correct an Inadequate Design Bases Calculation.

OTHER ACTIVITIES

4OA2 Problem Identification and Resolution

Main article: IP 71152

The inspectors reviewed condition reports associated with the selected components, operator actions, and operating experience notifications. There were issues with the resolution of a significant percentage of the condition reports reviewed by the inspectors.

However, because of the limited number of condition reports reviewed and the narrow focus of condition reports associated with the 10 components selected for the inspection, this is not a statistically significant evaluation of the licensees overall corrective action program. The inspectors noted that the licensee had difficulty tracking and trending items related to the environmental qualification program in their corrective action program based on the fact that condition reports known to the inspectors that had issues impacting or related to the environmental qualification program were not provided in the list of condition reports initially provided by the licensee.

4OA6 Meetings, Including Exit

Exit Meeting Summary

On August 10, 2017, the inspectors presented the inspection results to Mr. T. Evans, Acting Vice President, and other members of the licensee staff. The licensee acknowledged the issues presented. The licensee confirmed that any proprietary information reviewed by the inspectors had been returned or destroyed.

SUPPLEMENTAL INFORMATION

KEY POINTS OF CONTACT

Licensee Personnel

T. Evans, Vice President, Coordinator
R. Putnam, Director, Chief Engineer
V. Bacanskas, Director, Chief Engineer
P. Butler, Manager, Design and Program Engineering
J. Kirkpatrick, General Manager, Plant Operations
B. Lynch, Manager, Radiation Protection
S. Morris, Manager, Chemistry
R. Penfield, Director, Regulatory Assurance
G. Sullins, Senior Manager, Recovery
S. Pyle, Manager, Regulatory Assurance
M. Skartvedt, Manager, System Engineering
D. Vogt, Senior Manager, Operations
N. Mosher, Licensing Specialist, Regulatory Assurance

NRC Personnel

T. Farnholtz, Branch Chief, Engineering Branch 1
C. Henderson, Senior Resident
T. Sullivan, Resident Inspector
M. Tobin, Resident Inspector
J. Choate, Project Engineer

LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED

Opened and Closed

05000313/2017007-01 NCV Failure to Promptly Identify and Correct an Inadequate Design Bases Calculation (Section 1R21.2.2)

LIST OF DOCUMENTS REVIEWED

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