IR 05000247/2014002
| ML14132A170 | |
| Person / Time | |
|---|---|
| Site: | Indian Point  |
| Issue date: | 05/09/2014 |
| From: | Arthur Burritt Reactor Projects Branch 2 |
| To: | Ventosa J Entergy Nuclear Operations |
| Burrito A | |
| References | |
| IR-14-002 | |
| Download: ML14132A170 (48) | |
Text
{{#Wiki_filter: UNITED STATES May 9, 2014
SUBJECT:
INDIAN POINT POWER STATION - NRC INTEGRATED INSPECTION REPORT 05000247/2014002 AND 05000286/2014002
Dear Mr. Ventosa:
On March 31, 2014, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at your Indian Point Power Station, Units 2 and 3. The enclosed inspection report documents the inspection results, which were discussed on April 29, 2014, with you and other members of your staff.
The inspection 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.
The inspectors reviewed selected procedures and records, observed activities, and interviewed personnel.
This report documents four NRC-identified findings of NRC requirements, all of which were of very low safety significance (Green). These findings were determined to involve violations of NRC requirements. Additionally, one licensee-identified violation, which was determined to be of very low safety significance, is listed in this report. However, because of the very low safety significance and because they are entered into your corrective action program, the NRC is treating these findings as non-cited violations, consistent with Section 2.3.2.a of the NRC Enforcement Policy. If you contest the non-cited violations 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 Nuclear Regulatory Commission, ATTN: Document Control Desk, Washington, DC 20555-0001; with copies to the Regional Administrator, Region I; the Director, Office of Enforcement, United States Nuclear Regulatory Commission, Washington, DC 20555-0001; and the NRC Resident Inspector at Indian Point Power Station. In addition, if you disagree with the cross-cutting aspect assigned to any finding, or a finding not associated with a regulatory requirement in this report, you should provide a response with 30 days of the date of this inspection report, with the basis for your disagreement, to the Regional Administrator, Region I, and the NRC Resident Inspector at Indian Point Power Station.
Additionally, as we informed you in the most recent NRC integrated inspection report, cross-cutting aspects identified in the last six months of 2013 using the previous terminology were being converted in accordance with the cross-reference in Inspection Manual Chapter 0310. Section 4OA5 of the enclosed report documents the conversion of these cross-cutting aspects which will be evaluated for cross-cutting themes and potential substantive cross-cutting issues in accordance with Inspection Manual Chapter 0305 starting with the 2014 mid-cycle assessment review. If you disagree with the cross-cutting aspect assigned, you should provide a response within 30 days of the date of this inspection report, with the basis for your disagreement, to the Regional Administrator, Region I, and the NRC Resident Inspector at Indian Point Power Station.
In accordance with Title 10 of the Code of Federal Regulations (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 NRCs Public Document Room or from the Publicly Available Records component of the NRCs Agencywide Documents Access and Management System (ADAMS). ADAMS is accessible from the NRC website at http://www.nrc.gov/reading-rm/adams.html (the Public Electronic Reading Room).
Sincerely, /RA/ Arthur L. Burritt, Chief Reactor Projects Branch 2 Division of Reactor Projects Docket Nos. 50-247 and 50-286 License Nos. DPR-26 and DPR-64
Enclosure:
Inspection Report 05000247/2014002 and 05000286/2014002 w/Attachment: Supplementary Information
REGION I== Docket Nos. 50-247 and 50-286 License Nos. DPR-26 and DPR-64 Report Nos. 05000247/2014002 and 05000286/2014002 Licensee: Entergy Nuclear Northeast (Entergy) Facility: Indian Point Power Station, Units 2 and 3 Location: 450 Broadway, GSB Buchanan, NY 10511-0249 Dates: January 1, 2014 through March 31, 2014 Inspectors: J. Stewart, Senior Resident Inspector A. Patel, Resident Inspector G. Newman, Resident Inspector T. Lamb, Acting Resident Inspector E. Burket, Emergency Preparedness Inspector S. Chaudhary, Senior Engineering Inspector J. Furia, Senior Health Physicist E. H. Gray, Senior Reactor Inspector S. Pindale, Senior Reactor Inspector Approved By: Arthur L. Burritt, Chief Reactor Projects Branch 2 Division of Reactor Projects Enclosure
SUMMARY
IR 05000247/2014002, 05000286/2014002; 01/01/2014 - 03/31/2014; Indian Point Power
Station, Units 2 and 3; Maintenance Effectiveness; Maintenance Risk and Work Control; and Refueling and Other Outage Activities.
This report covered a three-month period of inspection by resident inspectors and announced inspections performed by regional inspectors. Inspectors identified four non-cited violations (NCVs) 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 (SDP), dated June 2, 2011. Cross-cutting aspects are determined using IMC 0310, Aspects Within the Cross-Cutting Areas, dated December 19, 2013. All violations of NRC requirements are dispositioned in accordance with the NRCs Enforcement Policy, dated July 9, 2013. The NRCs program for overseeing the safe operation of commercial nuclear power reactors is described in NUREG-1649, Reactor Oversight Process, Revision 5.
Cornerstone: Mitigating Systems
- Green.
The inspectors identified an NCV of Title 10 of the Code of Federal Regulations (10 CFR) Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, when Indian Point Energy Center (IPEC) staff did not evaluate spalled concrete in the Unit 3 service water pit ceiling slab to the extent required by Entergy procedures. Specifically, IPEC staff referenced an operability screening for a less significant spalled condition at this location that occurred in 2012, characterized spalls that exposed load carrying rebar as cosmetic, and did not consider the ongoing spalling. When identified by the inspectors to licensee staff, the licensee walked down the area, initiated condition report (CR)-IP3-2014-00405, and subsequently developed an operability determination and finite element analysis that determined the service water pit ceiling slab remained operable but degraded.
The failure of licensee staff to adequately perform an operability review of concrete degradation in the Unit 3 service water pit ceiling was contrary to self-imposed procedural standards and was within the licensee ability to foresee and correct and was a performance deficiency. The performance deficiency was determined to be more than minor because, if left uncorrected, it would have the potential to become a more significant safety concern.
Specifically, the failure to evaluate the spalling and exposed rebar in the operability screen resulted in IPEC staff not identifying the causes of ongoing spalling and scheduling corrective actions in a timeframe shown to be effective to maintain structural capability. The inspectors determined the finding could be evaluated using IMC 0609, Attachment 0609.04, and Initial Characterization of Findings. The inspectors screened the finding through IMC 0609, Appendix A, The Significance Determination Process for Findings At-Power, using Exhibit 2, Mitigating Systems Screening Questions. The finding screened as of very low safety significance (Green) because it did not result in the loss of operability or functionality.
The inspectors assigned a cross-cutting aspect in the area of Problem Identification and Resolution, Evaluation, because the licensee did not thoroughly evaluate the spalled condition and in completing the operability screening process, IPEC staff did not consider the additional spalled material that exposed rebar or causes of ongoing degradation when applying a prior operability screening for a previous less significant condition. [P.2] (Section 1R12)
- Green.
The inspectors identified an NCV of 10 CFR Part 50.65(a)(4) when Entergy did not adequately re-assess and manage risk when planned maintenance was not completed as scheduled. Specifically, IPEC staff did not re-assess the risk when the scheduled activity to remove pressurizer safety valves was delayed and did not inform the control room operators in the change in plant configuration due to the delayed maintenance activity. As a result, for about one shift, the control room operators were not aware of reactor coolant system (RCS)status (intact vs. not intact) and could have been challenged in the completing recovery actions in the event of loss of residual heat removal (RHR) cooling. This issue was entered into the licensees corrective action program as CR-IP2-2014-1986.
Not having re-assessed risk when safety valve removal was delayed and not keeping the control room operators aware of plant status due to the delayed maintenance activity resulted in the operators not knowing RCS status (intact vs. not intact) for about 8 hours, which was contrary to Entergys procedural requirements and was a condition reasonably within Entergys ability to foresee and correct and was a performance deficiency. The performance deficiency was more than minor because if left uncorrected, operator response to a loss of decay heat removal could lead to an incorrect decision which could adversely affect or delay recovery actions. The inspectors evaluated the finding using IMC 0609, Attachment 0609.04, Initial Characterization of Findings, which directed the inspectors to screen the finding through IMC 0609, Appendix G, Shutdown Operations, using Attachment 1, Checklist 2, PWR [pressurized-water reactor] Cold Shutdown Operation: Loops Filled and Inventory in Pressurizer. No deficiencies were identified in Checklist 2 which required a phase 2 or phase 3 quantitative assessment as the licensee maintained adequate mitigation capability. The inspectors concluded this finding had a cross-cutting aspect in the area of Human Performance, Work Management, when the licensee work process did not identify changing risk during removal of the pressurizer safety valves and manage the need for coordination between the work group and operations. Specifically, no controls were in place during the delay in pressurizer safety removal to ensure control room operators remained informed of the status of the reactor coolant system. The lack of coordination could have impacted operators ability to respond to a loss of RHR event. [H.5]
(Section 1R13)
- Green.
The Inspectors identified an NCV of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, when Entergy used a test procedure that was not appropriate to the circumstances and the operating loop of RHR was stopped during the conduct of the test. The test procedure did not assure technical specification (TS)requirements were met for an operating loop of RHR when steam generators were not available for backup decay heat removal. This issue was entered into the licensees corrective action program as CR-IP2-2014-2709.
The failure to accomplish testing using a procedure that ensured RCS loops were available for backup decay heat removal prior to stopping the operating RHR pump was a performance deficiency within the licensees ability to foresee and correct and should have been prevented. The finding was more than minor because if left uncorrected, would have the potential to become a more significant safety concern, specifically, a loss of decay heat removal cooling should the RHR pump fail to restart during the test without assurance that steam generators were available to remove decay heat. The inspectors evaluated the finding using IMC 0609, Attachment 0609.04, Initial Characterization of Findings, which directed the inspectors to screen the finding through IMC 0609, Appendix G, Shutdown Operations, using Attachment 1, Checklist 1, PWR Hot Shutdown Operation: Time to Core Boiling <2 Hours. No deficiencies were identified in Checklist 1 which required a phase 2 or phase 3 quantitative assessment as the licensee maintained adequate alternate mitigation capability and the finding screened to be of very low safety significance (Green). The inspectors concluded this finding had a cross-cutting aspect in the area of Human Performance, Design Margin, because the licensee did not put special attention in place to maintain safety-related equipment; specifically, when conducting testing that removed power from the running RHR loop without assurance that RCS loops remained filled and available for backup core cooling. [H.6] (Section 1R20)
Cornerstone: Barrier Integrity
- Green.
The inspectors identified an NCV of 10 CFR 50, Appendix B, Criterion V, when IPEC staff failed to follow fuel handling procedures which ensure that the correct spent fuel pool configuration is used in the development of the core offload plan, ensure that a cell location is visually verified as empty prior to loading, and ensure an evaluation is performed for any situation that results in a large or unexplained change in spent fuel handling machine (SFHM) load which resulted in two fuel assembly interference events in the Unit 2 spent fuel pool. This issue was entered into the licensees corrective action program (CAP) as CR-IP2-2014-1462.
This finding is more than minor as it represented a challenge to the human performance attribute of the Barrier Integrity cornerstone objective to provide reasonable assurance that physical design barriers (fuel cladding) protect the public from radionuclide releases caused by accidents or events. In accordance with IMC 0609, Significance Determination Process (SDP), Appendix A, The Significance Determination Process for Findings At-Power, Barrier Integrity Screening Questions, Section D, Spent Fuel Pool, the finding screened to be of very low safety significance (Green) when all screening questions were answered no. The event did not result in adverse impact to the decay heat removal capabilities of the spent fuel pool; the event did not result in detectible release of radionuclides; and the event did not result in the loss of spent fuel pool water inventory. The inspectors assigned a cross-cutting aspect in the Human Performance, Avoid Complacency, when the licensee staff failed to recognize and plan for the possibility of mistakes and failed to implement appropriate error reduction tools. [H.12] (Section 1R20)
Other Findings
A violation of very low safety significance that was identified by Entergy was reviewed by the inspectors. Corrective actions taken or planned by Entergy have been entered into Entergys corrective action program (CAP). This violation and corrective action tracking number are listed in Section 4OA7 of this report.
REPORT DETAILS
Summary of Plant Status
Unit 2 began the inspection period at 100 percent power and operated at full power until February 24 when the plant was shut down for a planned refueling and maintenance outage (2R21). Following refueling and maintenance activities, the reactor was started on March 18, 2014, and returned to power operation on March 19. Unit 2 achievedfull power on March 23 and remained at full power for the rest of the inspection period.
Unit 3 began the inspection period at 100 percent power. On January 6, 2014, Unit 3 tripped from full power due to a faulty feedwater valve controller that resulted in low steam generator level. The controller was repaired, and the unit restarted on January 8 and returned to full power on January 10. Unit 3 remained at full power for the remainder of the period.
REACTOR SAFETY
Cornerstones: Initiating Events, Mitigating Systems, and Barrier Integrity
1R01 Adverse Weather Protection
Readiness for Impending Adverse Weather Conditions
a. Inspection Scope
The inspectors reviewed Entergys preparations for the onset of cold weather on January 7, 2014. The inspectors reviewed the implementation of Indian Point adverse weather procedure OAP-48, Seasonal Weather Preparation, before the onset of and during this adverse weather condition. The inspectors walked down the listed plant areas to ensure system availability and that there were no problems as result of the severe weather. The inspectors verified that operator actions defined in Entergys adverse weather procedure maintained the readiness of essential systems. The inspectors discussed readiness and staff availability for adverse weather response with operations supervisors. The inspectors discussed cold weather preparedness with operators and maintained an awareness of cold weather issues throughout the cold weather periods. Documents reviewed for each section of this inspection report are listed in the Attachment.
Unit 2 emergency diesel generator (EDG) building Unit 2 turbine driven fire pump building Unit 3 feedwater regulating valve area and adjacent auxiliary feedwater room Unit 3 service water room
b. Findings
No findings were identified. ==1R04 Equipment Alignment
.1 Partial System Walkdowns
==
a. Inspection Scope
The inspectors performed partial walkdowns of the following systems: Unit 2 22 auxiliary boiler feedwater pump using licensee procedure 2-COL-21.3, Steam Generator Water Level, following failure of 21 auxiliary boiler feedwater pump to start from the alternate supply 12FD3 (CR-IP2-2014-364) on January 23, 2014 Backup spent fuel pit cooling system during core offload for refueling outage 2R21 on March 3, 2014 Unit 3 Auxiliary feedwater train alignment on January 17, 2014 31 and 33 EDGs and 480V switchgear room while the 32 EDG was tagged out for maintenance on January 21, 2014 Unit 2 appendix R diesel generator while the Unit 3 appendix R diesel generator was non-functional on March 24, 2014 The inspectors selected these systems based on their risk-significance relative to the reactor safety cornerstones at the time they were inspected. The inspectors reviewed applicable operating procedures, system diagrams, the Updated Final Safety Analysis Report (UFSAR), TSs, work orders, CRs, and the impact of ongoing work activities on redundant trains of equipment in order to identify conditions that could have impacted system performance of their intended safety functions. The inspectors also performed field walkdowns of accessible portions of the systems to verify system components and support equipment were aligned correctly and were operable. The inspectors examined the material condition of the components and observed operating parameters of equipment to verify that there were no unknown deficiencies. The inspectors also reviewed whether Entergy staff had properly identified equipment issues and entered them into the CAP for resolution with the appropriate significance characterization.
b. Findings
No findings were identified.
.2 Full System Walkdown
a. Inspection Scope
The inspectors performed a complete system walkdown of accessible portions of the Unit 2 safety injection system to verify the existing equipment lineup was correct. The inspectors reviewed operating procedures, surveillance tests, drawings, equipment line-up check-off lists, and the UFSAR to verify the system was aligned to perform its required safety functions. The inspectors also reviewed electrical power availability, component lubrication and equipment cooling, hanger and support functionality, and operability of support systems. The inspectors performed field walkdowns of accessible portions of the systems to verify system components and support equipment were aligned correctly and operable. The inspectors examined the material condition of the components and observed operating parameters of equipment to verify that there were no deficiencies. Additionally, the inspectors reviewed a sample of related CRs and work orders to ensure Entergy appropriately evaluated and resolved any deficiencies.
b. Findings
No findings were identified. ==1R05 Fire Protection