IR 05000354/2023010
| ML23215A014 | |
| Person / Time | |
|---|---|
| Site: | Hope Creek  |
| Issue date: | 08/03/2023 |
| From: | Brice Bickett Division of Operating Reactors |
| To: | Mcfeaters C Public Service Enterprise Group |
| References | |
| IR 2023010 | |
| Download: ML23215A014 (1) | |
Text
August 3, 2023
SUBJECT:
HOPE CREEK GENERATING STATION - BIENNIAL PROBLEM IDENTIFICATION AND RESOLUTION INSPECTION REPORT 05000354/2023010
Dear Charles McFeaters:
On July 11, 2023, the U.S. Nuclear Regulatory Commission (NRC) completed a problem identification and resolution inspection at your Hope Creek Generating Station (Hope Creek)
and discussed the results of this inspection with Robert DeNight, Site Vice President, and other members of your staff. The results of this inspection are documented in the enclosed report.
The NRC inspection team reviewed the stations problem identification and resolution program and the stations implementation of the program to evaluate its effectiveness in identifying, prioritizing, evaluating, and correcting problems, and to confirm that the station was complying with NRC regulations and licensee standards for problem identification and resolution programs.
Based on the samples reviewed, the team determined that your staffs performance in each of these areas adequately supported nuclear safety.
The team also evaluated the stations processes for use of industry and NRC operating experience information and the effectiveness of the stations audits and self-assessments.
Based on the samples reviewed, the team determined that your staffs performance in each of these areas adequately supported nuclear safety.
Finally, the team reviewed the stations programs to establish and maintain a safety conscious work environment, and interviewed station personnel to evaluate the effectiveness of these programs. Based on the teams observations and the results of these interviews the team found no evidence of challenges to your organizations safety conscious work environment. Your employees appeared willing to raise nuclear safety concerns through at least one of the several means available.
No findings or violations of more than minor significance were identified during this inspection. 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 (10 CFR) 2.390, Public Inspections, Exemptions, Requests for Withholding.
Sincerely, Brice A. Bickett, Chief Projects Branch 3 Division of Operating Reactor Safety
Docket No. 05000354 License No. NPF-57
Enclosure:
As stated
Inspection Report
Docket Number:
05000354
License Number:
Report Number:
Enterprise Identifier: I-2023-010-0014
Licensee:
Facility:
Hope Creek Generating Station
Location:
Hancocks Bridge, NJ
Inspection Dates:
June 12, 2023 to July 11, 2023
Inspectors:
S. Haney, Senior Resident Inspector
J. Brand, Reactor Inspector
B. Dyke, Operations Engineer
C. Khan, Senior Project Engineer
Approved By:
Brice A. Bickett, Chief
Projects Branch 3
Division of Operating Reactor Safety
SUMMARY
The U.S. Nuclear Regulatory Commission (NRC) continued monitoring the licensees performance by conducting a biennial problem identification and resolution inspection at Hope Creek Generating Station (Hope Creek), 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
No findings or violations of more than minor significance were identified.
Additional Tracking Items
None.
INSPECTION SCOPES
Inspections were conducted using the appropriate portions of the inspection procedures (IPs) in effect at the beginning of the inspection unless otherwise noted. Currently approved IPs with their attached revision histories are located on the public website at http://www.nrc.gov/reading-rm/doc-collections/insp-manual/inspection-procedure/index.html. Samples were declared complete when the IP requirements most appropriate to the inspection activity were met consistent with Inspection Manual Chapter (IMC) 2515, Light-Water Reactor Inspection Program - Operations Phase. The inspectors 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.
OTHER ACTIVITIES - BASELINE
71152B - Problem Identification and Resolution Biennial Team Inspection (IP Section 03.04)
- (1) The inspectors performed a biennial assessment of the effectiveness of PSEGs Problem Identification and Resolution program, use of operating experience, self-assessments and audits, and safety conscious work environment.
- Problem Identification and Resolution Program Effectiveness: The inspectors assessed the effectiveness of PSEGs Problem Identification and Resolution program in identifying, prioritizing, evaluating, and correcting problems. The inspectors also conducted a five-year review of the station service water system. The corrective actions for the following non-cited violations and findings were evaluated as part of the assessment: NCV 05000354/2021012-02, FIN 05000354/2021003-01, FIN 05000354/2021004-01, NCV 05000354/2022010-01, NCV 05000354/2022002-01, and NCV 05000354/2022002-03.
- Operating Experience: The inspectors assessed the effectiveness of PSEGs processes for use of operating experience.
- Self-Assessments and Audits: The inspectors assessed the effectiveness of PSEGs identification and correction of problems identified through audits and self-assessments.
- Safety Conscious Work Environment: The inspectors assessed the effectiveness of the stations programs to establish and maintain a safety conscious work environment.
INSPECTION RESULTS
Assessment 71152B Problem Identification and Resolution Program Effectiveness:
The inspectors determined that PSEGs problem identification and resolution program for Hope Creek was generally effective and adequately supported nuclear safety and security.
Identification: The team reviewed a sample of issues that have been processed through PSEG's problem identification and resolution program since the last biennial team inspection, including significant conditions adverse to quality, conditions adverse to quality, non-cited violations of regulatory requirements and other documented findings. The team determined that, in general, the station identified issues and entered them into the corrective action program at a low threshold.
Prioritization and Evaluation: Based on the samples reviewed, the team determined that, in general, PSEG appropriately prioritized and evaluated issues commensurate with the safety significance of the identified problem. In most cases, PSEG appropriately screened notifications (NOTFs) for operability and reportability, categorized notifications by significance, and assigned actions to the appropriate department for evaluation and resolution. However, the team identified one minor violation, regarding the evaluation of a degraded condition in the Hope Creek service water intake structure. The minor violation is documented below.
Corrective Action: The team determined that the overall problem identification and resolution program performance related to resolving problems was effective. In most cases, PSEG developed and implemented corrective actions to resolve problems in a timely manner.
However, the team identified one minor violation regarding a planned long term corrective action. The minor violation is documented below.
Assessment 71152B Operating Experience: The team determined that PSEG appropriately evaluated industry operating experience for its relevance to Hope Creek. In most cases, PSEG appropriately incorporated both external and internal operating experience into plant procedures and processes. However, the team identified one minor performance deficiency regarding an operating experience review performed in an equipment reliability evaluation. The minor performance deficiency is documented below.
Assessment 71152B Self-Assessment and Audits: The team reviewed a sample of self-assessments and audits, including self-assessments and audits of the PSEG corrective action and quality assurance programs to determine whether they appropriately assessed performance and identified areas for improvement. In most cases, the team concluded that PSEG had an effective self-assessments and audits process, and that the issues identified by those self-assessments and audits were addressed. However, the team identified one observation regarding a recommendation identified in the Focused Area Self-Assessment performed in preparation for this inspection. The observation is documented below.
Assessment 71152B Safety Conscious Work Environment: To determine whether underlying factors exist that would produce a reluctance to raise nuclear safety concerns, the team conducted interviews with 22 employees from a cross-section of the organization, including the Operations, Maintenance, Engineering, Security, Chemistry, Radiation Protection, and Emergency Preparedness departments. The team also interviewed the Employee Concerns Program managers and reviewed the Employee Concerns Program case log and select case files. The team also reviewed nuclear safety culture meeting minutes and the results of the most recently performed safety culture assessment. The team did not identify issues that represented challenges to the free flow of information, nor any underlying factors that could produce a reluctance to raise nuclear safety concerns. Based on inspection interviews and insights obtained from safety culture and other relevant assessments, the conditions at Hope Creek were conducive to a Safety Conscious Working Environment.
Minor Performance Deficiency 71152B Minor Performance Deficiency: Equipment reliability evaluation (Order 70224588), was conducted to review elevated C' station service water pump vibration levels. The equipment reliability evaluation determined with high confidence that the stuffing box shaft sleeve chrome oxide coating degraded, and the chipped, delaminated coating created foreign material that entered the shaft bearing. Step 4.10.1 of LS-AA-125-1001, Cause Analysis, states To be most effective, PERFORM Operating Experience (OE) reviews twice: once prior to cause determination and again after causes have been identified. The equipment reliability evaluation conducted broad searches on the terms Elevated Vibrations, large vertical pump, service water pump. The equipment reliability evaluation did not perform more focused operating experience searches after the cause had been identified.
As a result, relevant operating experience was not documented in the equipment reliability evaluation or recognized by the evaluation team. For example, the Generic Equivalent Replacement Change Package performed in 1999 that accepted the chrome oxide coating for application at Hope Creek documented that Salem and Millstone station experience has shown chrome oxide coating separates from base metal when used as a wear surface at the stuffing box location. Additionally, the NRC issued NCV 05000311/2018-002-01, "Inadequate Design Change for Service Water Pumps" in NRC Inspection Report 05000311/2018002 (ML18207A221) for chromium oxide delamination in the Salem 26 service water pump for which Salem performed an equipment reliability evaluation (Order 70199758). Further, Salem performed item equivalency evaluation (Order 80124207) in 2019 to accept Ultimet laser cladding for use at Salem.
Screening: The inspectors determined the performance deficiency was minor. This issue is captured in NOTF 20939922. This performance deficiency was evaluated in accordance with the guidance in Inspection Manual Chapter (IMC) 0612, Appendix B, Issue Screening, and Appendix E, Examples of Minor Issues. The inspectors determined this issue was not of more than minor significance because the condition would have been corrected. Specifically, engineers identified cracking on the 'C' station service water pump shaft sleeve coating in December 2021 and wrote NOTF 20893348 prior to any elevated vibration conditions on the pump. The Equipment Reliability Action Tracking Items process (N2 NOTF 20893609 and Order 70221150) drove the review of the shaft sleeve coating change independently, prior to the performance of the equipment reliability evaluation.
Minor Violation 71152B Minor Violation: Equipment reliability evaluation (Order 70224588), "'C' Station Service Water Pump Elevated Vibration Levels," was conducted by PSEG to evaluate a high vibration condition in the 'C' station service water pump resulting from delamination of a chrome oxide coating on the pump's stuffing box shaft sleeve. Long term corrective action 70224588-0140 is planned to replace the chrome oxide coated stuffing box shaft sleeve chrome oxide coating with upgraded Ultimet laser cladding on all four station service water pumps. Engineering evaluated the change to the new Ultimet laser cladding (Order 70221150) and incorrectly determined that no item equivalency evaluation was necessary.
Screening: The inspectors determined the performance deficiency was minor. This issue is captured in NOTF 20939922. This performance deficiency was evaluated in accordance with the guidance in Inspection Manual Chapter (IMC) 0612, Appendix B, Issue Screening, and Appendix E, Examples of Minor Issues. The inspectors determined this issue was not of more than minor significance because it did not adversely affect any cornerstone objectives since the shaft sleeves with Ultimet laser cladding have not yet been installed in the Hope Creek station service water system.
Enforcement:
10 CFR Part 50, Appendix B, Criterion III, Design Control, requires that measures be established for the selection and review of materials for suitability of application.
PSEG procedure CC-AA-300, "Procurement Engineering Support Activities," Revision 5, Step 4.1.1, Item 3, requires the station to perform an item equivalency evaluation to document the acceptability of non-identical replacement items by evaluating form, fit, and function. VCC-AA-3000, "Standard Item Equivalency," Revision 0, defines Form as, "The physical characteristics, material composition, design ratings, dimensions, code applicability, QA requirements, and NRC requirements associated with an item."
Contrary to the above, on May 19, 2022, PSEG evaluated the material composition change of the service water stuffing box shaft sleeve coating from chrome oxide to Ultimet laser cladding and incorrectly determined an item equivalency evaluation was not required.
This failure to comply with 10 CFR Part 50, Appendix B, Criterion III, constitutes a minor violation that is not subject to enforcement action in accordance with the NRCs Enforcement Policy.
Minor Violation 71152B Minor Violation: During a walkdown of the 'A' and 'C' station service water pump room in the service water intake structure, the inspectors identified that the support frame of H1EA -10-C-514, Service Water Pump Lube Oil Control Panel, was severely corroded. The support of the control panel is subjected to wetted conditions from pump packing and strainer leakage. The panel is nonsafety-related equipment which houses spray wash flow indication components, but is in close proximity to the safety-related, Seismic Category I, 'A' and 'C' Service Water Strainers (1AF-509 and 1CF-509), and is therefore designed to meet Seismic II/I requirements. A Seismic II/I interaction is a condition that could potentially reduce the capability of a Seismic Category I structure, system or component to perform its safety-related function as a result of the structural failure of an adjacent non-Category I structure, system or component during a safe shutdown earthquake. DE-PZ.ZZ-0011, "Seismic II/I Program," Section 2.0.C states that, Structures, systems and components in the proximity of safety-related items shall be designed and constructed so as to not fail during [a safe shutdown earthquake] in such a way that they cause the safety-related items to fail to perform their function.
A number of notifications have documented this condition dating back to 2012 and as recently as 2022. A seismic and structural evaluation of the panel's degraded condition was last performed in 2014. NOTF 20897820 was written in 2022, stating that continued degradation could result in the potential loss of the panel's Seismic II/I design function. Operations performed a functionality assessment of the condition and determined that the 'A' station service water system remained operable because the spray wash flow indication is not required for system operability. OP-AA-108-115, "Operability Determination & Functionality Assessments," Step 4.4.2.5 directs that an operability determination should include the potential effect of the degraded condition on the affected structure, system or component's ability to perform specified safety functions. Step 4.4.3.1 notes that plant specific design basis events should be considered when performing operability determinations. While the functionality assessment of notification 20897820 evaluated the loss of the spray wash flow indications and the impact of their loss on the station service water system, a review of the Seismic II/I interaction and potential loss of the panel's Seismic II/I design function was not performed.
PSEG conducted engineering walkdowns in response to the inspectors' questions confirmed the support frame had further degraded from corrosion since its last evaluation performed in 2014, and determined that the anchor bolts cannot perform their function. PSEG performed technical evaluation (Order 70230018) and Operable with Engineering Justification and subsequently determined that panel 10-C-514 is structurally adequate in its current condition to meet Seismic II/I requirements and not impact any nearby safety-related equipment.
Screening: The inspectors determined the performance deficiency was minor. This issue is captured in NOTF 20939945. This performance deficiency was evaluated in accordance with the guidance in Inspection Manual Chapter (IMC) 0612, Appendix B, Issue Screening, and Appendix E, Examples of Minor Issues. The inspectors determined this issue was not of more than minor significance because it did not adversely affect any cornerstone objectives since the panel remained structurally adequate to meet Seismic II/I requirements.
Enforcement:
10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, requires, in part, that activities affecting quality shall be prescribed by documented instructions, procedures, or drawings of a type appropriate to the circumstances and shall be accomplished in accordance with these instructions, procedures, or drawings.
PSEG programmatic standard DE-PZ.ZZ-0011, "Seismic II/I Program," Section 12.0 states in part that the Quality Assurance Program describes the requirements to assure the control of activities affecting the safety-related function of structures, systems or components at Hope Creek Generating Station as required by 10 CFR 50, Appendix B, and that those requirements shall apply to activities related to the Seismic II/I program.
OP-AA-108-115, "Operability Determination & Functionality Assessments," Step 4.4.2.5 directs that an operability determination should include the potential effect of the degraded condition on the affected structure, system or component's ability to perform specified safety functions. Step 4.4.3.1 notes that plant specific design basis events should be considered when performing operability determinations.
Contrary to the above, on February 25, 2022, the functionality assessment of notification 20897820 documenting service water control panel support channel corrosion did not evaluate the Seismic II/I interaction or the potential loss of the panel's Seismic II/I design function.
This failure to comply with 10 CFR Part 50, Appendix B, Criterion V, constitutes a minor violation that is not subject to enforcement action in accordance with the NRCs Enforcement Policy.
Observation: Self-Assessment Safety Conscious Work Environment Policy Recommendation Not Addressed 71152B As a part of the Focused Area Self-Assessment (Order 70227854) performed in March in preparation for this inspection, a pulsing survey was sent to gather additional safety conscious work environment insights. The responses identified that 100 percent of the respondents indicated that they are willing to raise safety concerns and that they were also all aware that PSEG Nuclear has a safety conscious work environment policy. There was some variation however, identified in the responses on what the safety conscious work environment policy is. The Focused Area Self-Assessment subsequently recommended that the fleet reinforce the safety conscious work environment policy, LS-AA-3, "Safety Conscious Work Environment," via a Nuclear Communication. This recommendation was not captured as an action in the report or as an assignment in the order, and no Nuclear Communication was issued to the fleet.
The interviews conducted with PSEG personnel by the team during this inspection confirmed that site personnel are willing to raise safety concerns and identified a similar variation in responses regarding the safety conscious work environment policy. The Focused Area Self-Assessment recommendation was a missed opportunity to reinforce the safety conscious work environment policy at the fleet level. This issue is captured in NOTF 20939007, and communications regarding the PSEG safety conscious work environment policy were distributed in the Hope Creek Weekly Alignment Package on June 19, 2023 and in a Nuclear Communication on June 22,
EXIT MEETINGS AND DEBRIEFS
The inspectors verified no proprietary information was retained or documented in this report.
- On July 11, 2023, the inspectors presented the biennial problem identification and resolution inspection results to Robert DeNight, Site Vice President, and other members of the licensee staff.
DOCUMENTS REVIEWED
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
Corrective Action
Documents
20398027
20577490
20616574
20644637
20666626
20781588
20781588
20867692
20877063
20878950
20879321
20880247
20880344
20885221
20897820
20897899
20898182
20898243
20899016
20907120
20907137
20909712
20910533
20910533
20920455
20922956
20924488
20931449
20936699
Corrective Action
Documents
Resulting from
20937867
20939006
20939007
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
Inspection
20939182
20939183
20939184
20939185
20939187
20939190
20939191
20939192
20939193
20939194
20939195
20939196
20939197
20939201
20939336
20939339
20939341
20939428
20939494
20939679
20939804
20939922
20939934
20939935
20939945
20939960
20939974
20939975
20940451
20940471
20940506
Drawings
M-10-1, Sheet 1
Hope Creek Generating Station, Service Water
Drawings
M-10-1, Sheet 2
Hope Creek Generating Station, Service Water
Drawings
M-10-1, Sheet 3
Hope Creek Generating Station, Service Water
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
Drawings
M-10-1, Sheet 4
Hope Creek Generating Station, Service Water
Engineering
Changes
Generic
Equivalent
Replacement
Change Package
Equivalency Replacement HC SW Shaft Sleeve Chrome-
plating
05/14/1999
Miscellaneous
10855-G-052
Seismic II/I Evaluation Program for the Hope Creek
Generating Station
Miscellaneous
DE-PZ.ZZ-0011
Seismic II/I Program
Procedures
Procurement Engineering Support Activities
Procedures
Item Equivalency Evaluations
Procedures
Engineering Technical Evaluations
Procedures
Transient Loads
Procedures
Control of Transient Combustible Material
Procedures
HC.ER-PS.FP-
0001-A5
Programmatic Standard for Fire Protection
Procedures
HC.IC-GP.ZZ-
0115
Transmitter Isolation/Restoration Procedure Sensitive Rack
Instrumentation, Instrument Rack 10C005 - RPV Channel C
Procedures
Station Event-Free Clock Program
Procedures
NRC Inspection Preparation and Response
Procedures
Operating Experience Program
Procedures
Issue Identification and Screening Process
Procedures
Issue Identification and Screening Process
Procedures
Corrective Action Program
Procedures
Cause Analysis
Procedures
Safety Conscious Work Environment
Procedures
Nuclear Safety Culture Monitoring
Procedures
Technical Fundamental Skills
Procedures
Conduct of Troubleshooting
Procedures
Conduct of Troubleshooting
Procedures
Maintenance Alterations Process
Procedures
Integrated Risk Management
Procedures
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
Procedures
Procedures
OP-AA-108-115-
1002
Supplemental Considerations for On-Shift Operability
Determinations
Procedures
PIA-018
Root Cause Evaluation Template
Procedures
PIA-035
Cause Analysis Manual
Procedures
PIA-036
Equipment Reliability Evaluation (ERE)
Procedures
VCC-AA-3000
Standard Item Equivalency Process, NISP-EN-02
Procedures
Work Activity Risk Management
Work Orders
30185282
30340883
50196833
235036
236755
237633
238434
240257
242039
60060811
60087495
60096784
60105959
60128469
60150259
70054662
70064658
70198964
70199758
218377
218387
219712
220690
220691
221150
222139
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
223906
224525
224588
226516
226871
227209
227854
224461
230018
80118982
80124207
80134488