IR 05000354/2017007

Design Bases Assurance (Environmental Qualification Program) Inspection Report 05000354/2017007

ML17188A450
Person / Time
Site:	Hope Creek
Issue date:	07/07/2017
From:	Glenn Dentel Engineering Region 1 Branch 2
To:	Sena P Public Service Enterprise Group
Dentel G
References
IR 2017007 IR 2017007
Download: ML17188A450 (16)

Text

uly 7, 2017

SUBJECT:

HOPE CREEK GENERATING STATION UNIT 1 - DESIGN BASES ASSURANCE (ENVIRONMENTAL QUALIFICATION PROGRAM) INSPECTION REPORT 05000354/2017007

Dear Mr. Sena:

On July 6, 2017, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at the Hope Creek Generating Station Unit (HCGS). The enclosed inspection report documents the inspection results, which were discussed on July 6, 2017, with Mr. Edward Casulli, Plant Manager, and other members of your staff.

NRC inspectors examined activities conducted under your license as they relate to safety and compliance with the Commissions rules and regulations and with the conditions of your license.

In conducting the inspection, the team examined Public Service Enterprise Group Nuclear (PSEG) implementation of the electrical equipment environmental qualification program required by Title 10, Code of Federal Regulations (10 CFR) 50.49 for maintaining the qualified status of equipment during the life of the plant. The inspection involved field walkdowns, examination of selected procedures, calculations and records, and interviews with station personnel.

This report documents one NRC-identified finding of very low safety significance (Green). The finding was determined to be a violation of NRC requirements. However, because of the very low safety significance and because it was entered into your corrective action program, the NRC is treating this finding as a non-cited violation (NCV) consistent with Section 2.3.2.a of the NRCs Enforcement Policy. If you contest the NCV in this 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 I; the Director, Office of Enforcement, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001; and the NRC Resident Inspector at HCGS. In addition, if you disagree with the cross-cutting aspect assignment 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 I, and the NRC Resident Inspector at HCGS. In accordance with 10 CFR 2.390 of the NRCs Rules of Practice, a copy of this letter, its enclosure, and your response (if any) will be available electronically for public inspection in the NRC Public Docket Room or from the Publicly Available Records component of NRCs document system, Agencywide Documents Access and Management System (ADAMS).

ADAMS is accessible from the NRC Web site at http://www.nrc.gov/reading-rm/adams.html (the Public Electronic Reading Room).

Sincerely,

/RA/

Glenn Dentel, Chief Engineering Branch 2 Division of Reactor Safety Docket No. 50-354 License No. NPF-57

Enclosure:

Inspection Report 05000354/2017007 w/Attachment: Supplemental Information

REGION I==

Docket No: 50-354 License No: NPF-57 Report No: 05000354/2017007 Licensee: PSEG Nuclear LLC Facility: Hope Creek Generating Station Unit 1 Location: P.O. Box 236, Hancocks Bridge, NJ 08038 Inspection Period: May 8 through May 26, 2017 Inspectors: K. Mangan, Senior Reactor Inspector, Division of Reactor Safety (DRS),

Team Leader J. Brand, Reactor Inspector, DRS M. Orr, Reactor Inspector, DRS Approved By: Glenn Dentel, Chief Engineering Branch 2 Division of Reactor Safety Enclosure

SUMMARY OF FINDINGS

IR 05000354/2017007; 5/8/2017 - 7/6/2017; Hope Creek Generating Station Unit 1; Design

Bases Assurance Inspection (Programs).

The report covers the Design Bases Assurance Inspection - Programs conducted by a team of three U.S. Nuclear Regulatory Commission (NRC) inspectors. The inspectors identified one non-cited violation (NCV), which was of very low safety significance (Green). The significance of most findings is indicated by their color (i.e., greater than Green, or Green, White, Yellow,

Red) and determined using Inspection Manual Chapter (IMC) 0609, Significance Determination Process, dated April 29, 2015. Cross-cutting aspects are determined using IMC 0310,

Aspects Within Cross-Cutting Areas, dated December 4, 2014. All violations of NRC requirements are dispositioned in accordance with the NRCs Enforcement Policy, dated November 1, 2016. The NRCs program for overseeing the safe operation of commercial nuclear power reactors is described in NUREG-1649, Reactor Oversight Process, Revision 6, dated July 2016.

Cornerstone: Barrier Integrity

Green.

The team identified a Green, non-cited violation (NCV) of Technical Specification (TS)3.6.5.1, for failure to maintain secondary containment integrity. Specifically, while Hope Creek station was operating in mode 1, PSEG personnel did not ensure secondary containment door R-4302 was properly latched (dogged) closed in accordance with procedure CC-AA-201, Plant Barrier Control Program. The licensees failure to ensure the door was properly dogged closed was a performance deficiency and resulted in a degraded secondary containment barrier for approximately 44 hours5.092593e-4 days <br />0.0122 hours <br />7.275132e-5 weeks <br />1.6742e-5 months <br />. The team determined that PSEG operated in violation of the TS LCO which requires restoration of secondary containment integrity within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> or be in at least hot shutdown within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in cold shutdown within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. Following identification of the door condition by the team PSEG personnel properly dogged the door closed restoring secondary containment.

This finding was determined to be of more than minor significance because it is associated with the configuration control attribute of the Barrier Integrity cornerstone and affected the cornerstone objective of providing reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. Specifically, in its un-dogged position the door would not have remained closed, as required to maintain secondary containment integrity, during all design basis accidents. Using IMC 0609, Appendix A, The Significance Determination Process for Findings At-Power, Exhibit 3, Barrier Integrity Screening Questions, the team determined that this finding was of very low safety significance (Green) because the finding did not represent an actual open pathway in the physical integrity of primary reactor containment (valves, airlocks, etc.), containment isolation system (logic and instrumentation), and heat removal components. The finding was determined to be associated with the cross-cutting area of Human Performance - Procedure Adherence (H.8), in that, licensee personnel did not follow process, procedures, and work instructions which required the secondary containment door to be closed and dogged. (Section 4OA2)

REPORT DETAILS

REACTOR SAFETY

Cornerstones: Initiating Events, Mitigating Systems, and Barrier Integrity

1R21 Design Bases Assurance Inspection (Programs) (IP 71111.21N - 9 samples)

.1 Inspection Sample Selection Process

The inspection team assessed the implementation of PSEGs Environmental Qualification (EQ) program, established to meet the requirements of Title 10 of the Code of Federal Regulations (10 CFR) 50.49, Environmental Qualification of Electrical Equipment Important to Safety for Nuclear Power Plants. The inspection team performed an inspection as outlined in NRC Inspection Procedure (IP) 71111.21N, 1, Environmental Qualification under 10 CFR 50.49 Programs, Processes, and Procedures. The team reviewed components that were either safety-related equipment relied upon to remain functional during and following design basis events, non-safety-related components whose failure could prevent safety-related equipment from performing design functions, and certain post-accident monitoring equipment. The team then determined which components environment would be adversely affected by postulated post-accident environmental conditions (temperature, pressure, radiation level or flood level) and reviewed information contained in the HCGS Probabilistic Risk Assessment (PRA) and the U.S. NRCs Standardized Plant Analysis Risk (SPAR)model for HCGS to determine risk significant components that were also required to meet EQ requirements. Additionally, the team interviewed plant staff, reviewed design records, and discussed the EQ program with the resident inspectors to assist in the selection of components. Finally, the team ensured that different types of components were selected included pump motors, motor-operated valves, solenoid valves, limit switches, radiation monitors, and flow/level transmitters that were located both inside and outside of primary containment. Based on these reviews, the team selected nine EQ components and associated sub-components (seals, cables, connectors, and lubricants) for inspection.

.2 Results of Detailed Reviews (9 Samples)

a. Inspection Scope

The inspection was conducted as outlined in NRC IP 71111.21N. The team assessed PSEGs implementation of the EQ program required by 10 CFR 50.49. The team reviewed EQ program-related procedures, component EQ files, EQ test records, equipment maintenance and operating history, maintenance and operating procedures, vendor documents, design documents, previously identified deficiencies, and design calculations. The team also interviewed plant staff knowledgeable of the design, maintenance, and operation of the selected components. The review and associated interviews were performed to evaluate whether PSEGs staff properly maintained the equipment qualifications for electrical equipment important to safety through plant life (repair, replacement, modification, and plant life extension);established and maintained required EQ documentation records; and implemented an effective corrective action program to identify and correct EQ-related deficiencies and evaluate EQ-related industry operating experience. The team also performed walkdowns (where accessible) of selected components to verify equipment was installed as described in HCGS environmental qualification component documentation files, the environmental conditions were consistent with those assumed in the evaluations; whether equipment surrounding EQ component could fail in a manner that would prevent the components safety function from being performed; and that the components were installed in their tested configuration. The component and associated subcomponents inspected were:

Residual heat removal injection valve (BC-HV-F017B)

Actuator, motor, limit switches, cable, and breaker High pressure coolant injection inboard steam isolation valve (FD-HV-F002)

Actuator, motor, limit switches, cable, and breaker Reactor building 480V motor control center (PH-10-B-222)

MCC, breakers, and supply cable Residual heat removal pump room unit cooler (1AVH210)

Fan motor, cable, and temperature instrumentation Reactor vessel level instrument (1BBLT-3622B)

Differential pressure instrument, and cable Safety relief valve (PSV F013F)

Solenoids, acoustic monitor, and cable Filter recirculation and ventilation system radiation monitor (SP-1SPXE-4811)

Radiation monitoring instrument Emergency core cooling system pressure transmitter (BB-1BBPT-N094B-B21)

Differential pressure instrument, and cable Safety auxiliary cooling system heat exchanger bypass valve (EG-1EGHS-2457A)

Temperature control, actuator, solenoid, and switches In addition to the inspection of the selected components, the team performed general plant walkdowns to determine whether components located in areas susceptible to a high-energy line break were properly evaluated for operation in a harsh environment.

The team also reviewed procurement records and inspected a sample of replacement parts stored in the warehouse to verify EQ parts approved for installation in the plant were properly identified and controlled; and that storage time and environmental conditions did not adversely affect the components qualified life or service life. Finally, the team reviewed a sample of components that had been removed from the EQ program to determine if PSEG had correctly determined that the components no longer were required to meet 10 CFR 50.49. Documents reviewed for this inspection are listed in the Attachment.

b. Findings

No findings were identified.

4OA2 Identification and Resolution of Problems (IP 71152)

a. Inspection Scope

The team reviewed a sample of problems that PSEG had previously identified and entered into the corrective action program. The team reviewed a sample of these issues to verify an appropriate threshold for identifying issues and to evaluate the effectiveness of corrective actions. Additionally, the team evaluated whether deficiencies identified during the inspection were properly documented and evaluated in the corrective action program.

b. Findings

Introduction:

The team identified a Green, NCV of TS 3.6.5.1, because PSEG did not maintain secondary containment integrity. Specifically, while HCGS was operating in mode 1, the team identified that secondary containment door R-4302 was not properly latched (dogged) closed in accordance with procedure CC-AA-201, Plant Barrier Control Program. Because the door was not dogged, secondary containment was determined to be inoperable.

Description:

During a walkdown of the reactor building on May 10, 2017, the team identified that secondary containment door R-4302 was closed but not properly latched (un-dogged). PSEG personnel accompanying the inspectors immediately contacted the control room and were authorized to properly dog the door closed. Operators verified that negative secondary containment pressure was being maintained and that there were no control room alarms associated with the door. An extent of condition walkdown of accessible areas was also performed by operators to verify there were no other similar conditions.

The team reviewed procedure CC-AA-201 plant barrier control program and noted that Section 3.3 of the procedure requires that doors should be placed in their required position and secured properly after use in order to be considered capable of performing their credited barrier functions. A door latch or closure mechanism may be required for the door to meet its design function. For water tight doors to be considered secured, all dogs must be engaged unless a different number and configuration were previously evaluated by engineering as meeting the design function.

The team found that the design functions of Door 4302, include:

Secondary Containment Boundary - to prevent the release of radionuclides following design basis event (DBE) and to ensure that safety-related and non-safety related reactor building exhaust fans can maintain a slightly negative pressure within secondary containment rooms.

Internal Flooding Barrier - to ensure that design basis pipe failures do not cause flooding in adjacent safety related areas.

Fire Protection Barrier - to ensure that fire in adjacent areas do not spread.

Following this review the team questioned whether the door in an un-dogged position was operable for each of the credited functions. PSEG performed a prompt technical assessment to address these questions. PSEG determined that pre-event secondary containment ventilation was not impacted based on a review of secondary containment vacuum levels; internal flooding assessment for safety related equipment was not impacted by the door not being secured and the door in the closed position was adequate to address the very low combustible material loading for the area. However, the licensee determined that the door may not remain closed during certain DBE in order to maintain secondary containment integrity. PSEG determined that if the door opened following a seismic event or as result of the pressure transient in secondary containment following a large-break-loss-of-coolant accident, secondary containment would not be maintained at a negative pressure and an open pathway to the environment from secondary containment would exist making secondary containment inoperable. As a result operators would be required to manually close and dog the door to restore secondary containment integrity.

The team reviewed TS LCO 3.6.5.1, Secondary Containment Integrity, and determined that it requires that secondary containment integrity be maintained operable while in Mode 1. The LCO requires restoration of secondary containment integrity within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> or be in at least hot shutdown within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in cold shutdown within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. The team concluded that because the door was un-dogged for 44 hours5.092593e-4 days <br />0.0122 hours <br />7.275132e-5 weeks <br />1.6742e-5 months <br /> PSEG operated in violation of the TS LCO. Specifically, because PSEG personnel did not identify or recognize that secondary containment door R-4302 was not properly dogged closed, the TS specified actions were not implemented within the required time of the LCO.

Analysis:

The licensees failure to ensure secondary containment Door 4302 was properly dogged closed in accordance with procedure CC-AA-201, Plant Barrier Control Program, was a performance deficiency and resulted in an in-operable secondary containment barrier for approximately 44 hours5.092593e-4 days <br />0.0122 hours <br />7.275132e-5 weeks <br />1.6742e-5 months <br />, which is greater than the TS allowed outage time of 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

This finding was determined to be of more than minor significance because it is associated with the configuration control attribute of the Barrier Integrity cornerstone and affected the cornerstone objective of providing reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events.

Specifically, in its un-dogged position the door would not have remained closed, as required to maintain secondary containment integrity, during all design basis accidents.

Using IMC 0609, Appendix A, The Significance Determination Process for Findings At-Power, Exhibit 3, Barrier Integrity Screening Questions, the team determined that this finding was of very low safety significance (Green) because the finding did not represent an actual open pathway in the physical integrity of primary reactor containment (valves, airlocks, etc.), containment isolation system (logic and instrumentation), or heat removal components. The finding was determined to be associated with the cross-cutting area of Human Performance - Procedure Adherence (H.8), in that, licensee personnel did not follow process, procedures, and work instructions which required the secondary containment door to be closed and dogged.

Enforcement:

TS 3.6.5.1 requires that secondary containment integrity be maintained while operating in mode 1. The TS requires restoration of secondary containment integrity within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> or be in at least hot shutdown within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in cold shutdown within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. Contrary to the above, on May 8, 2017, PSEG personnel did not to ensure secondary containment door R-4302 was properly latched (dogged) closed in accordance with procedure CC-AA-201, Plant Barrier Control Program, which resulted in a degraded secondary containment barrier for approximately 44 hours5.092593e-4 days <br />0.0122 hours <br />7.275132e-5 weeks <br />1.6742e-5 months <br /> and required TS actions were not taken. The TS LCO was restored when PSEG personnel properly dogged the door closed on May 10, 2017. Because the licensee entered the issue into their corrective action program (notification 20764632)and the finding is of very low safety significance (Green), this violation is being treated as an NCV, consistent with the Section 2.3.2 of the NRC Enforcement Policy (NCV 05000354/2017007-01, Secondary Containment Integrity Not Maintained Due to Door not Properly Dogged).

4OA6 Meetings, including Exit

On May 26, 2017, the team presented the inspection results to Mr. Steve Poorman, Operations Director, and other members of the PSEG staff. The team reviewed proprietary information, which was returned to PSEG at the end of the inspection.

On July 6, 2017, the team conducted a phone exit to discuss the inspection results with Mr. Edward Casulli, Plant Manager, and other members of the PSEG staff. The team verified that no proprietary information was documented in the report.

ATTACHMENT

SUPPLEMENTAL INFORMATION

KEY POINTS OF CONTACT

PSEG Personnel

E. Casulli, Plant Manager

A. Ford, Design Engineer

C. Lukacsy, Operations Supervisor

T. MacEwen, Compliance Engineer

A. Ochoa, Compliance Engineer

M. Richers, Engineering Manager

A. Bhuta, Design Engineer

S. Poorman, Director, Operations

D. Mannai, Director, Regulatory Compliance

F. Powell, Superintendent Inventory & Logistics Nuclear Procurement Operations

V. Cavallaro, Nuclear Equipment Operator

S. Todd, Nuclear Equipment Operator

LIST OF ITEMS OPENED, CLOSED, DISCUSSED, AND UPDATED

Opened/Closed

05000354/2017007-01 NCV Secondary Containment Integrity Not Maintained Due to Door not Properly Dogged