IR 05000271/2015001
| ML15044A023 | |
| Person / Time | |
|---|---|
| Site: | Vermont Yankee  |
| Issue date: | 02/12/2015 |
| From: | Raymond Mckinley NRC/RGN-I/DRP/PB5 |
| To: | Wamser C Entergy Nuclear Operations |
| McKinley R | |
| References | |
| IR 2015001 | |
| Download: ML15044A023 (15) | |
Text
ary 12, 2015
SUBJECT:
VERMONT YANKEE NUCLEAR POWER STATION - NRC INTEGRATED INSPECTION REPORT 05000271/2015001
Dear Mr. Wamser:
On January 24, 2015, the U.S. Nuclear Regulatory Commission (NRC) completed its final inspection under the Operating Reactor Assessment Program at your Vermont Yankee Nuclear Power Station (Vermont Yankee). The enclosed inspection report documents the inspection results, which were discussed on January 29, 2015, 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 one violation of NRC requirements, which was of very low safety significance (Green). 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, consistent with Section 2.3.2.a of the NRC Enforcement Policy.
On January 12, 2015, Vermont Yankee certified the permanent removal of fuel from the reactor vessel (Agencywide Documents Access and Management System (ADAMS) Accession No.
ML15013A426). On January 20, 2015, the NRC notified Vermont Yankee that the Operating Reactor Assessment Program would cease on January 24, 2015, and that implementation of the Decommissioning Power Reactor Inspection Program would begin on January 25, 2015 (ADAMS Accession No. ML15020A482).
If you contest the non-cited violation 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 Vermont Yankee. In addition, if you disagree with the cross-cutting aspect assigned to the finding 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 Regional Administrator, Region I, and the NRC Resident Inspector at Vermont Yankee. 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 ADAMS. ADAMS is accessible from the NRC website at http://www.nrc.gov/reading-rm/adams.html (the Public Electronic Reading Room).
Sincerely,
/RA/
Raymond R. McKinley, Chief Reactor Projects Branch 5 Division of Reactor Projects Docket No. 50-271 License No. DPR-28
Enclosure:
Inspection Report 05000271/2015001 w/Attachment: Supplementary Information
REGION I==
Docket No. 50-271 License No. DPR-28 Report No. 05000271/2015001 Licensee: Entergy Nuclear Operations, Inc. (Entergy)
Facility: Vermont Yankee Nuclear Power Station Location: Vernon, VT 05354 Dates: January 1 through January 24, 2015 Inspectors: S. Rutenkroger, PhD, Senior Resident Inspector, Division of Reactor Projects (DRP)
S. Rich, Resident Inspector, DRP Approved by: Raymond R. McKinley, Chief Reactor Projects Branch 5 Division of Reactor Projects Enclosure
SUMMARY
IR 05000271/2015001; 01/01/2015 - 01/24/2015; Vermont Yankee Nuclear Power Station (Vermont Yankee); Surveillance Testing.
This report covered a one-month period of inspection by resident inspectors. Inspectors identified one finding of very low safety significance (Green), which was a non-cited violation (NCV). 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 19, 2013. Cross-cutting aspects are determined using IMC 0310, Aspects Within the Cross-Cutting Areas, dated December 4, 2014. 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: Initiating Events
- Green.
The inspectors identified an NCV of Title 10 of the Code of Federal Regulations (10 CFR) 50, Appendix B, Criterion III, because Entergy did not establish measures to assure the design basis was correctly translated into procedures and instructions.
Specifically, an incorrect assumption in calculation VYC-332 led to reduced water level in the reactor vessel while using the shutdown cooling system. Entergys immediate corrective actions included entering the issue into their corrective action program as condition report CR-VTY-2015-0045.
The inspectors determined that Entergys failure to establish measures to assure the design basis was correctly translated into procedures and instructions in accordance with 10 CFR 50, Appendix B, Criterion III was a performance deficiency that was within Entergys ability to foresee and correct, and should have been prevented. This finding is more than minor because it impacted the configuration control attribute of the Initiating Events cornerstone and its objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during shutdown. Specifically, an incorrect assumption in calculation VYC-332 led to reduced water level in the reactor vessel while using the shutdown cooling system, increasing the likelihood of temperature stratification. In accordance with IMC 0609.04, Initial Characterization of Findings, and Exhibit 2 of IMC 0609, Appendix G, Attachment 1, Shutdown Operations Significance Determination Process Phase 1, issued May 9, 2014, the inspectors determined that this finding is of very low safety significance (Green). Although the performance deficiency resulted in a reduced water level, there was no loss of inventory event and the calibration error would not prevent operators from detecting decreasing vessel level.
Additionally, the reduced water level did not actually affect the ability of the RHR system to perform shutdown cooling. This finding has a cross-cutting aspect in Human Performance, Challenge the Unknown, because Entergy did not challenge the uncertainty of the conflicting vessel level indications. [H.11] (Section 1R22)
REPORT DETAILS
Summary of Plant Status
Vermont Yankee was in a shutdown status at the beginning of the inspection period. On January 12, 2015, Vermont Yankee certified the permanent removal of fuel from the reactor vessel (Agencywide Documents Access and Management System (ADAMS) Accession No.
ML15013A426). On January 20, the NRC notified Vermont Yankee that the Operating Reactor Assessment Program would cease on January 24, 2015, and that implementation of the Decommissioning Power Reactor Inspection Program would begin on January 25, 2015 (ADAMS Accession No. ML15020A482).
The inspectors also verified that all the fuel was safely removed from the reactor vessel and stored in the spent fuel pool.
REACTOR SAFETY
Cornerstones: Initiating Events, Mitigating Systems, Barrier Integrity
1R04 Equipment Alignment
Partial System Walkdowns (71111.04Q - 1 sample)
a. Inspection Scope
The inspectors performed partial walkdowns of the A and B emergency diesel generators and support systems while the startup transformers were out of service on January 12. 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 and system diagrams and the impact of ongoing work activities on related 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 deficiencies.
Documents reviewed for each section of this inspection report are listed in the
.
b. Findings
No findings were identified.
1R13 Maintenance Risk Assessments and Emergent Work Control
a. Inspection Scope
The inspectors reviewed station evaluation and management of plant risk during the activities listed below to verify that Entergy performed the appropriate risk assessments prior to removing equipment for work. The inspectors selected these activities based on potential risk significance relative to the reactor safety cornerstones. The inspectors verified that Entergy personnel performed risk assessments as required by 10 CFR 50.65(a)(4) and that the assessments were accurate and complete. The inspectors reviewed the scope of maintenance work and discussed the results of the assessment with the outage control center staff and the site probabilistic safety assessment engineer to verify plant conditions were consistent with the risk assessment. The inspectors also reviewed the technical specification requirements and inspected portions of redundant safety systems, when applicable, to verify risk analysis assumptions were valid and applicable requirements were met.
Vessel head removal the week of January 6 A emergency diesel generator surveillance run with the startup transformers out of service the week of January 19
b. Findings
No findings were identified.
1R18 Plant Modifications
.1 Temporary Modifications
a. Inspection Scope
The inspectors reviewed the temporary modification to the source range monitors (SRMs) implemented by engineering change (EC) 54805, Install jumpers to bypass the SRM downscale rod withdrawal blocks during defueling, to determine whether the modification affected the safety functions of systems that are important to safety. The inspectors reviewed 10 CFR 50.59 documentation and post-modification testing results, and conducted field walkdowns of the modifications to verify that the temporary modifications did not degrade the design bases, licensing bases, and performance capability of the source range monitors and rod withdrawal blocks. The inspectors also interviewed engineering and operations personnel regarding the modification.
b. Findings
No findings were identified.
.2 Permanent Modifications
a. Inspection Scope
The inspectors evaluated a voltage tap change to the startup transformers by EC 50013.
The inspectors verified that the design bases, licensing bases, and performance capability of the affected systems were not degraded by the modification. In addition, the inspectors reviewed modification documents associated with the upgrade and design change, including the revised calculation of the system voltage under the minimum grid voltage and maximum load conditions expected during the SAFSTOR period.
b. Findings
No findings were identified.
1R20 Refueling and Other Outage Activities
a. Inspection Scope
During the defueling outage, the inspectors monitored controls associated with the following outage activities:
Configuration management, including maintenance of defense-in-depth, commensurate with the outage plan for the key safety functions and compliance with the applicable technical specifications when taking equipment out of service Implementation of clearance activities and confirmation that tags were properly hung and that equipment was appropriately configured to safely support the associated work and/or testing Installation and configuration of reactor coolant pressure, level, and temperature instruments to provide accurate indication and instrument error accounting Status and configuration of electrical systems and switchyard activities Monitoring of decay heat removal operations Impact of outage work on the ability of the operators to operate the spent fuel pool cooling system Reactor water inventory controls, including flow paths, configurations, alternative means for inventory additions, and controls to prevent inventory loss Activities that could affect reactivity Maintenance of secondary containment as required by the technical specifications Defueling activities, including fuel handling Fatigue management Identification and resolution of problems related to refueling outage activities The resident inspectors verified that all the fuel was permanently removed from the reactor vessel and placed in the spent fuel pool on January 12, 2015.
b. Findings
No findings were identified.
1R22 Surveillance Testing
a. Inspection Scope
The inspectors reviewed test data from the wide range level instrument (LT-2-3-70)calibration on December 30, 2014, to verify test results satisfied technical specifications, the Updated Final Safety Analysis Report, and Entergys procedure requirements. The inspectors assessed whether the test acceptance criteria were clear, tests demonstrated operational readiness and were consistent with design documentation, test instrumentation had current calibrations and the range and accuracy for the application, tests were performed as written, and applicable test prerequisites were satisfied. Upon test completion, the inspectors considered whether the test results supported that equipment was capable of performing the required safety functions.
b. Findings
Introduction.
The inspectors identified a Green NCV of 10 CFR 50, Appendix B, Criterion III, because Entergy did not establish measures to assure the design basis was correctly translated into procedures and instructions. Specifically, an incorrect assumption in calculation VYC-332 led to reduced water level in the reactor vessel while using the shutdown cooling system.
Description.
On December 29, 2014, operators shut down and commenced cooling down Vermont Yankee to begin the defueling outage. Because the density of water changes with temperature, the operators need to use different instruments to monitor the level of water in the vessel during cold shutdown than they use during power operation.
Two instruments are typically used during the initial period of cold shutdown while the vessel head is still installed, the head-on shutdown level instrument and the wide range level instrument. The head-on shutdown level instrument is expected to be unreliable until the reference leg is filled, and the wide range level instrument requires recalibration for the cold conditions before it reads accurately.
On December 30, technicians recalibrated the wide range level instrument for cold conditions using procedure OP 5371, Instrument and Control Routine Outage Activities. Later on December 30, technicians refilled the reference leg for the head-on shutdown level instrument and found that it indicated a much lower level than the wide range instrument. They performed troubleshooting on both instruments and were unable to determine the cause of the discrepancy. On December 31, operators raised reactor vessel level above the main steam lines and were able to confirm by the rate of level increase that the head-on shutdown level instrument was accurate. The operators and technicians attributed the incorrect level indication on the wide range level instrument to the slight vacuum present in the vessel with the head on.
During the initial cold shutdown period, vessel level had been approximately 150. This is low enough to be on-scale for the reactor protection system level instruments, which are calibrated for hot conditions. The upper limit for these instruments is less than 180 at hot conditions, so if vessel level is within the intended band of 185-195 during cold conditions, the instruments are off-scale high. They were not, and the operators did not recognize this as an indication of low level, nor did they question the validity of the calibration on the wide range level instrument.
Procedure OPOP-RHR-2124, Residual Heat Removal, contains the directions for running the RHR system in the shutdown cooling mode. It directs that vessel level be maintained greater than 185 in order to reduce the likelihood of reactor water temperature stratification. If unrecognized and uncorrected, reactor water temperature stratification can cause rapid temperature changes on pump restart, which could challenge the technical specification limits on vessel cooldown. The inspectors reviewed data recorded by the plant computer and determined that temperature stratification had not occurred during the time of lowered level.
The calibration criteria in OP 5371 were calculated in VYC-332, Reactor Water Level and Pressure Head and dP, dated May 25, 2000. The inspectors identified that the calculation for the cold shutdown setpoints for the wide range level instrument assumed that the temperature of the water in the reactor vessel was 212°F, which resulted in using a low density for water in the calculation. Actual reactor vessel water temperature was approximately 100°F when the instrument was in use, and the lowest procedurally-allowable temperature was 80°F. Because of the increased density, the weight of the water that corresponded to the calibrated level of 185 was actually a significantly lower level. The inspectors informed Entergy of the incorrect assumption, and they wrote condition report CR-VTY-2015-0045 to document it.
Analysis.
The inspectors determined that Entergys failure to establish measures to assure the design basis was correctly translated into procedures and instructions in accordance with 10 CFR 50, Appendix B, Criterion III was a performance deficiency that was within Entergys ability to foresee and correct, and should have been prevented.
This finding is more than minor because it impacted the configuration control attribute of the Initiating Events cornerstone and its objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during shutdown. Specifically, an incorrect assumption in calculation VYC-332 led to reduced water level in the reactor vessel while using the shutdown cooling system, increasing the likelihood of temperature stratification.
In accordance with IMC 0609.04, Initial Characterization of Findings, and Exhibit 2 of IMC 0609, Appendix G, Attachment 1, Shutdown Operations Significance Determination Process Phase 1, issued May 9, 2014, the inspectors determined that this finding is of very low safety significance (Green). Although the performance deficiency resulted in a reduced water level, there was no loss of inventory event and the calibration error would not prevent operators from detecting decreasing vessel level. Additionally, the reduced water level did not actually affect the ability of the RHR system to perform shutdown cooling.
This finding has a cross-cutting aspect in Human Performance, Challenge the Unknown, because Entergy did not challenge the uncertainty of the conflicting vessel level indications. [H.11]
Enforcement.
10 CFR 50, Appendix B, Criterion III, Design Control, states, in part, that measures shall be established to assure that the design basis, as defined in § 50.2 and as specified in the license application, are correctly translated into procedures and instructions. Contrary to the above, on May 25, 2000, Entergy approved a calculation for the calibration setpoints of the wide range level instrument that included a non-conservative assumption. Entergys immediate corrective actions included entering the issue into their corrective action program as condition report CR-VTY-2015-0045.
Because this violation was of very low safety significance (Green) and Entergy entered this issue into their corrective action program, this violation is being treated as an NCV, consistent with Section 2.3.2 of the Enforcement Policy. (NCV 050000271/2015001-01, Incorrect Calculation Assumption Led to Reduced Reactor Vessel Level While in Cold Shutdown)
OTHER ACTIVITIES
4OA2 Problem Identification and Resolution
Routine Review of Problem Identification and Resolution Activities
a. Inspection Scope
As required by Inspection Procedure 71152, Problem Identification and Resolution, the inspectors routinely reviewed issues during baseline inspection activities and plant status reviews to verify that Entergy entered issues into their corrective action program at an appropriate threshold, gave adequate attention to timely corrective actions, and identified and addressed adverse trends. In order to assist with the identification of repetitive equipment failures and specific human performance issues for follow-up, the inspectors performed a daily screening of items entered into the corrective action program and periodically attended condition report review group meetings.
b. Findings
No findings were identified.
4OA6 Meetings, Including Exit
On January 29, the inspectors presented the inspection results to Mr. Christopher Wamser, Site Vice President, and other members of the Entergy staff who acknowledged the inspection results. The inspectors verified that no proprietary information was retained by the inspectors or documented in this report.
ATTACHMENT:
SUPPLEMENTARY INFORMATION
KEY POINTS OF CONTACT
Vermont Yankee Personnel
- C. Wamser, Site Vice President
- M. Romeo, Decommissioning Plant Manager
- J. Boyle, Engineering and Technical Director
- P. Paradis, Decommissioning Director
- E. Harms, Operations Manager
- R. Felumb, Performance Improvement Manager
- P. Ryan, Security Manager
- D. Tkatch, Radiation Protection Manager
- C. Chappell, Licensing and CA&A Manager
- C. Tabone, Control Room Supervisor
- J. Rogers, Design Engineering Manager
- J. Laughney, QA Supervisor
- R. Congdon, Shift Manager
- J. Garozzo, Senior Engineer
LIST OF ITEMS OPENED, CLOSED, DISCUSSED AND UPDATED
Opened/Closed
- 05000271/2015001-01 NCV Incorrect Calculation Assumption Led to Reduced Reactor Vessel Level While in Cold Shutdown