ML042080530
| ML042080530 | |
| Person / Time | |
|---|---|
| Site: | Vermont Yankee File:NorthStar Vermont Yankee icon.png |
| Issue date: | 07/26/2004 |
| From: | Anderson C NRC/RGN-I/DRP/PB5 |
| To: | Thayer J Entergy Nuclear Operations |
| References | |
| FOIA/PA-2005-0031 IR-04-003 | |
| Download: ML042080530 (36) | |
See also: IR 05000271/2004003
Text
July 26, 2004
Mr. Jay K. Thayer
Site Vice President
Entergy Nuclear Operations, Inc.
Vermont Yankee Nuclear Power Station
P.O. Box 0500
185 Old Ferry Road
Brattleboro, VT 05302-0500
SUBJECT:
VERMONT YANKEE NUCLEAR POWER STATION - NRC INTEGRATED
INSPECTION REPORT 05000271/2004003
Dear Mr. Thayer:
On June 30, 2004, the US Nuclear Regulatory Commission (NRC) completed an inspection at
your Vermont Yankee Nuclear Power Station (VY). The enclosed report documents the
inspection findings which were discussed on July 12, 2004, with 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 finding of very low safety significance (Green) which was also
determined to involve a violation of NRC requirements. Because of the very low safety
significance and because the finding was entered into your corrective actions program, the
NRC is treating it as a non-cited violation (NCV), consistent with Section VI.A of the NRCs
Enforcement Policy. If you contest this non-cited violation, 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, D.C. 20555-0001; with
copies to the Regional Administrator Region I; the Director, Office of Enforcement, United
States Nuclear Regulatory Commission, Washington, D.C. 20555-0001; and the NRC Resident
Inspector at the Vermont Yankee Nuclear Power Station.
Jay K. Thayer
2
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 Document Room or from the Publicly Available Records (PARS) component of
NRCs document 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/
Clifford J. Anderson, Chief
Projects Branch 5
Division of Reactor Projects
Docket No.
50-271
License No.
Enclosure:
Inspection Report 05000271/2004003
w/Attachment: Supplemental Information
Docket No. 50-271
License No. DPR-28
Jay K. Thayer
3
cc w/encl:
M. R. Kansler, President, Entergy Nuclear Operations, Inc.
G. J. Taylor, Chief Executive Officer, Entergy Operations
J. T. Herron, Senior Vice President and Chief Operating Officer
D. L. Pace, Vice President, Engineering
B. OGrady, Vice President, Operations Support
J. M. DeVincentis, Manager, Licensing, Vermont Yankee Nuclear Power Station
Operating Experience Coordinator - Vermont Yankee Nuclear Power Station
J. F. McCann, Director, Nuclear Safety Assurance
M. J. Colomb, Director of Oversight, Entergy Nuclear Operations, Inc.
J. M. Fulton, Assistant General Counsel, Entergy Nuclear Operations, Inc.
S. Lousteau, Treasury Department, Entergy Services, Inc.
Administrator, Bureau of Radiological Health, State of New Hampshire
Chief, Safety Unit, Office of the Attorney General, Commonwealth of Mass.
D. R. Lewis, Esquire, Shaw, Pittman, Potts & Trowbridge
G. D. Bisbee, Esquire, Deputy Attorney General, Environmental Protection
Bureau
J. Block, Esquire
D. Katz, Citizens Awareness Network (CAN)
M. Daley, New England Coalition on Nuclear Pollution, Inc. (NECNP)
R. Shadis, New England Coalition Staff
C. McCombs, Commonwealth of Massachusetts, SLO Designee
G. Sachs, President/Staff Person, c/o Stopthesale
J. Sniezek, PWR SRC Consultant
J. P. Matteau, Executive Director, Windham Regional Commission
State of New Hampshire, SLO Designee
State of Vermont, SLO Designee
Jay K. Thayer
4
Distribution w/encl:
H. Miller, RA/J. Wiggins, DRA (1)
C. Anderson, DRP
D. Florek, DRP
D. Pelton, Senior Resident Inspector
C. Miller, RI EDO Coordinator
J. Clifford, NRR
Region I Docket Room (with concurrences)
DOCUMENT NAME:C:\\ORPCheckout\\FileNET\\ML042080530.wpd
After declaring this document An Official Agency Record it will/will not be released to the Public.
To receive a copy of this document, indicate in the box: "C" = Copy without attachment/enclosure "E" = Copy with attachment/enclosure "N" = No copy
OFFICE RI:DRP
RI:DRP
RI:DRP
NAME
Pelton/CJA for
Florek/CJA for
Anderson/CJA
DATE
07/26/04
07/26/04
07/26/04
OFFICIAL RECORD COPY
Enclosure
i
U.S. NUCLEAR REGULATORY COMMISSION
REGION I
Docket No.
50-271
Licensee No.
Report No.
Licensee:
Entergy Nuclear Vermont Yankee, LLC
Facility:
Vermont Yankee Nuclear Power Station
Location:
320 Governor Hunt Road
Vernon, Vermont
05354-9766
Dates:
April 1, 2004 - June 30, 2004
Inspectors:
David L. Pelton, Senior Resident Inspector
Beth E. Sienel, Resident Inspector
E. Harold Gray, Senior Reactor Inspector
Todd J. Jackson, Senior Project Engineer
James D. Noggle, Senior Health Physicist
Larry L. Scholl, Senior Reactor Inspector
Keith A. Young, Senior Reactor Inspector
Amar C. Patel, Reactor Inspector
Jennifer A. Bobiak, Reactor Inspector
Thomas P. Sicola, Reactor Inspector
Approved by:
Clifford J. Anderson, Chief
Projects Branch 5
Division of Reactor Projects
Enclosure
ii
TABLE OF CONTENTS
SUMMARY OF FINDINGS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . iii
REACTOR SAFETY . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1
1R01
Adverse Weather . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1
1R02
Evaluations of Changes, Tests, or Experiments . . . . . . . . . . . . . . . . . . . . . . . . . 1
1R04
Equipment Alignments . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2
1R05
Fire Protection
. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3
1R06
Flood Protection Measures . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3
1R08
Inservice Inspection . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4
1R11
Licensed Operator Requalification . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5
1R12
Maintenance Effectiveness . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6
1R13
Maintenance Risk Assessment and Emergent Work Evaluation . . . . . . . . . . . . 6
1R14
Personnel Performance During Non-routine Plant Evolutions . . . . . . . . . . . . . . 7
1R15
Operability Evaluations
. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8
1R16
Operator Workarounds . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8
1R17
Permanent Plant Modifications . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9
1R19
Post Maintenance Testing . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 10
1R20
Refueling and Outage Activities . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 10
1R22
Surveillance Testing . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15
1R23
Temporary Modifications . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 16
1EP6
Drill Evaluation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 16
RADIATION SAFETY . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 17
2OS1 Access Control to Radiologically Significant Areas . . . . . . . . . . . . . . . . . . . . . 17
2OS2 ALARA Planning and Controls . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 17
OTHER ACTIVITIES (OA) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 18
4OA1 Performance Indicator Verification . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 18
4OA2 Identification and Resolution of Problems . . . . . . . . . . . . . . . . . . . . . . . . . . . . 18
4OA3 Event Followup
. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 19
4OA5 Other Activities . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 20
4OA6 Meetings, including Exit . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 20
SUPPLEMENTAL INFORMATION . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . A-1
KEY POINTS OF CONTACT . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . A-1
LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED . . . . . . . . . . . . . . . . . . . . . . . . . . . A-1
LIST OF DOCUMENTS REVIEWED
. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . A-2
LIST OF ACRONYMS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . A-6
Enclosure
iii
SUMMARY OF FINDINGS
IR 05000271/2004003; 04/01/04 - 06/30/04; Vermont Yankee Nuclear Power Station; Refueling
and Outage Activities.
This report covered a 13-week period of baseline inspection conducted by resident inspectors.
Additionally, announced inspections were performed by regional inspectors in the areas of
occupational radiation protection; evaluations of changes, tests, and experiments; in-service
inspections; and permanent plant modifications. One Green non-cited violation (NCV) was
identified. The significance of most findings is indicated by their color (Green, White, Yellow,
Red) using Inspection Manual Chapter (IMC) 0609, Significance Determination Process
(SDP). Findings for which the SDP does not apply may be Green or be assigned a severity
level after NRC management review. The NRC's program for overseeing the safe operation of
commercial nuclear power reactors is described in NUREG-1649, "Reactor Oversight Process,"
Revision 3, dated July 2000.
A.
NRC-Identified and Self-Revealing Findings
Cornerstone: Barrier Integrity
(Green) A self-revealing, non-cited violation (NCV) of 10 CFR 50 Criterion XVI was
identified in that Entergy personnel did not develop effective corrective actions to
prevent recurrence following a 2001 event wherein control room operators did not verify
a suction path existed prior to starting the residual heat removal (RHR) system pump
being used to support shutdown cooling (SDC) operations which caused the pump to
trip. On April 10, 2004, an identical event occurred and again resulted in a trip of the
RHR pump being used to support SDC operations.
The finding is greater than minor since it is associated with the Fuel Cladding
Configuration Control Attribute of the Barrier Integrity Cornerstone and because it
affects the associated Cornerstone objective. The inspectors conducted a SDP Phase 1
screening of the finding in accordance with IMC 0609, Appendix G, Shutdown
Operations Significance Determination Process [SDP]. In accordance with the SDP,
the inspectors determined that the finding was of very low safety significance (Green)
since the RHR pump was restarted within 15 minutes of being tripped and an adequate
SDC thermal margin was maintained as demonstrated by a calculated reactor coolant
system (RCS) time-to-boil of greater than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
A contributing cause of this finding is related to the Cross-Cutting area of Problem
Identification and Resolution. As stated above, Entergy personnel did not develop
effective corrective actions to prevent recurrence following a 2001 event wherein control
room operators did not verify a suction path existed prior to starting the RHR system
pump being used to support SDC operations which caused the pump to trip. Entergys
corrective actions relied on the operators skill to verify a suction path was open prior to
restarting the RHR pump rather than proceduralize the step. As a result, an identical
event occurred in April 2004 again resulting in a trip of the RHR pump being used to
support SDC operations. (Section 4OA3.1)
Summary of Findings (contd)
Enclosure
iv
B.
Licensee Identified Findings
None.
Enclosure
REPORT DETAILS
Summary of Plant Status
Vermont Yankee Nuclear Power Station entered the inspection period at or near full power.
The reactor was shutdown on April 3, 2004, in support of planned refueling outage (RFO) 24.
Reactor startup activities began on May 3, 2004, following the completion of RFO 24. The
reactor was returned to full power operation on May 8, 2004. On June 18, 2004, an automatic reactor scram occurred as a result of a turbine trip following multiple faults-to-ground on the 22
kilovolt (KV) electrical system. The reactor remained shutdown for the rest of the inspection
period.
1.
REACTOR SAFETY
Cornerstones: Initiating Events, Mitigating Systems, Barrier Integrity
1R01
Adverse Weather (71111.01)
a.
Inspection Scope (one sample)
The inspectors reviewed measures established by Entergy for the restoration from cold
weather operations. The inspectors reviewed Vermont Yankee Operating Procedure
(OP) 2196, Preparations for Cold Weather Operations, Form VYOPF 2196.02, Cold
Weather Restoration Operations Checklist, discussed the completion of items with
operations personnel to confirm the items on the checklist had been completed or were
appropriately tracked for completion, and independently walked down portions of the
plant to verify selected actions to restore from cold weather operations had been
completed appropriately.
b.
Findings
No findings of significance were identified.
1R02
Evaluations of Changes, Tests, or Experiments (71111.02)
a.
Inspection Scope (eight samples)
The inspectors reviewed the 10 CFR 50.59 safety evaluations or screening evaluations
associated with plant modifications being installed during the current refueling outage to
support a proposed power uprate. The inspectors assessed the adequacy of the safety
evaluations through interviews with the cognizant plant staff and review of supporting
documentation to verify the changes were performed in accordance with 10 CFR 50.59
and when required, NRC approval was obtained prior to implementation. The inspectors
also reviewed a sample of changes the licensee had evaluated (using a screening
process) and determined to be outside of the scope of 10 CFR 50.59, therefore not
requiring a full safety evaluation. The inspectors performed this review to determine if
Entergy conclusions with respect to 10 CFR 50.59 applicability were appropriate. A
listing of the modifications for which associated safety evaluations, safety evaluation
2
Enclosure
screenings, and other documents were reviewed is provided in the Attachment to this
report.
b.
Findings
No findings of significance were identified.
1R04
Equipment Alignments
1.
Complete Equipment Alignment (71111.04S)
a.
Inspection Scope (one sample)
The inspectors performed a complete equipment alignment inspection of the accessible
portions of the core spray (CS) system. The inspectors walked down the CS system,
both inside and outside of the primary containment, and compared actual equipment
alignment to approved piping and instrumentation diagrams, operating procedure
lineups, the Vermont Yankee updated final safety analysis report (UFSAR), and the
Vermont Yankee design basis document (DBD). The inspectors observed valve
positions, the availability of power supplies, and the general condition of selected
components to verify there were no unidentified deficiencies. The inspectors also
confirmed that licensee-identified equipment problems had been entered into the
corrective actions program.
b.
Findings
No findings of significance were identified.
2.
Partial Equipment Alignments (71111.04)
a.
Inspection Scope (four samples)
The inspectors performed four partial system walkdowns of risk significant systems to
verify system alignment and to identify any discrepancies that would impact system
operability. Observed plant conditions were compared with the standby alignment of
equipment specified in the licensees system operating procedures and drawings. The
inspectors also observed valve positions, the availability of power supplies, and the
general condition of selected components to verify there were no obvious deficiencies.
The inspectors verified the alignment of the following systems:
The spent fuel pool (SFP) cooling system while the A train of the residual heat
removal (RHR) system was unavailable to support shutdown cooling on June 6,
2004;
The B train of the standby gas treatment (SBGT) system during planned
maintenance on the A SBGT fan on June 7, 2004;
3
Enclosure
The A train of SBGT during planned instrument calibrations on the B train of
SBGT on June 8; and
The emergency diesel generators (EDGs), start-up transformers, the diesel oil
storage tank (DOST) following the main transformer fire on June 18, 2004.
b.
Findings
No findings of significance were identified.
1R05
Fire Protection (71111.05Q)
a.
Inspection Scope (nine samples)
The inspectors identified fire areas important to plant risk based on a review of Entergys
the Vermont Yankee Safe Shutdown Capability Analysis, the Fire Hazards Analysis, and
the individual plant evaluation of external events (IPEEE). The inspectors toured plant
areas important to safety in order to verify the suitability of Entergys control of transient
combustibles and ignition sources, and the material condition and operational status of
fire protection systems, equipment, and barriers. The following fire areas were
inspected:
Reactor building, 252 foot elevation-S1 cable trays (CFZ-3/4);
Reactor building, 252 foot elevation-S2 cable trays (CFZ-3/4);
Reactor building, 252 foot elevation, North (FZ RB3);
Reactor building, 252 foot elevation, South (FZ RB4);
Reactor building, 280 foot elevation, Recirc MG set area (SZ RB-MG);
Turbine building, all elevations (FA TB);
Torus room, 213 foot elevation, North (FZ RB1);
Torus room, 213 foot elevation, South (FZ RB2);
345 KV relay house.
b.
Findings
No findings of significance were identified.
1R06
Flood Protection Measures (71111.06)
a.
Inspection Scope (one sample)
The inspectors reviewed Entergys established flood protection barriers and procedures
for coping with internal flooding in the EDG rooms including Vermont Yankee Off-
Normal Procedure (ON) 3148, Loss of Service Water; and ON 3158, Reactor Building
High Area Temperature/Water Level. The inspectors reviewed internal flooding
information contained in Entergys IPEEE, in the UFSAR, and in the Internal Flooding
DBD as it related to the EDG rooms. Finally, the inspectors performed walk-downs of
flood vulnerable portions of the EDG rooms to ensure equipment and structures needed
4
Enclosure
to mitigate an internal flooding event were as described in the IPEEE and the DBD.
Additionally, the inspectors reviewed condition reports (CRs) related to internal flooding
and the EDG rooms to ensure identified problems were properly addressed for
resolution.
b.
Findings
No findings of significance were identified.
1R08
Inservice Inspection (71111.08G)
a.
Inspection Scope (four samples)
The inspectors assessed the inservice inspection (ISI) activities using the criteria
specified in the American Society of Mechanical Engineers (ASME) Boiler and Pressure
Vessel Code,Section XI.
The inspectors observed selected in-process non-destructive examination (NDE)
activities, reviewed documentation and interviewed personnel to verify that the activities
were performed in accordance with the ASME Boiler and Pressure Vessel Code Section
XI requirements. The sample selection was based on the inspection procedure
objectives and risk priority of those components and systems where degradation would
result in a significant increase in risk of core damage. The inspectors reviewed a
sample of condition reports and quality assurance audit reports to assess the licensees
effectiveness in problem identification and resolution. The specific ISI activities selected
for review included:
Observation of the ultrasonic testing (UT) manual technique, UT procedure, weld
overlay calibration test block, and performance of pre and post examination
calibration for UT of the CS system N5A safe-end to nozzle structural weld
overlay;
Review of the computer based UT procedure and observation of its application
for the reactor vessel welds and the eddy current (ET) examination method to
quantify clad crack shadowing of volumetric vessel weld examinations and the
results for the reactor vessel flange-to-vessel weld;
Observation of the UT examination of a pre-existing reactor vessel weld
indication for verification that the indication was appropriately characterized and
had not increased in dimension since the previous examination;
Review of CS system sparger video-visual examination records;
Review of the inspection scope expansion and disposition of two small linear
indications on a standby liquid control system socket weld (SL11-F12); and
Review of the reactor vessel internals project (BWRVIP-03 Rev 6) procedure
and observation of some of the initial visual examinations.
In response to Entergys extended power up-rate request and recent industry operating
experience, the inspectors observed portions of the steam dryer visual testing (VT) type
5
Enclosure
1 and type 3 examinations and reviewed the documented examination reports. The
examination reports documented that cracks were identified on both the internal and
external surfaces of the steam dryer. The inspectors reviewed Entergys corrective
actions for these indications to ensure that the actions were appropriate. Specifically,
the inspectors reviewed the weld repair activities for the two cracks identified on the
external surface of the steam dryer. The inspectors also reviewed the vendor technical
reports which justified operation for the next operating cycle at the current maximum
licensed power level without repair of the indications identified on internal portions of the
steam dryer.
b.
Findings
No findings of significance were identified.
1R11
Licensed Operator Requalification (71111.11Q)
a.
Inspection Scope (one sample)
The inspectors observed simulator examinations for one operating crew to assess the
performance of the licensed operators and the ability of Entergys Training Department
staff to evaluate licensed operator performance. Operating crew performance was
evaluated during a simulated main steam line break inside the drywell coincident with a
loss of normal power. The inspectors evaluated the crews performance in the areas of:
Clarity and formality of communications;
Ability to take timely actions;
Prioritization, interpretation, and verification of alarms;
Procedure use;
Control board manipulations;
Oversight and direction from supervisors; and
Group dynamics.
Crew performance in these areas was compared to Entergy management expectations
and guidelines as presented in the following documents:
Vermont Yankee Administrative Procedure (AP) 0151, Responsibilities and
Authorities of Operations Department Personnel;
AP 0153, Operations Department Communication and Log Maintenance; and
Vermont Yankee Department Procedure (DP) 0166, Operations Department
Standards.
The inspectors verified that the crew completed the critical tasks listed in the associated
simulator evaluation guide (SEG). The inspectors also compared simulator
configurations with actual control board configurations. For any weaknesses identified,
the inspectors observed the licensee evaluators to verify that they also noted the issues
to be discussed with the crew.
6
Enclosure
b.
Findings
No findings of significance were identified.
1R12
Maintenance Effectiveness (71111.12Q)
a.
Inspection Scope (three samples)
The inspectors performed three issue/problem-oriented inspections of actions taken by
Entergy in response to the following issues:
As-found local leakage rate testing (LLRT) failures of the high pressure coolant
injection (HPCI) turbine exhaust vacuum breakers;
Repeat failures of the C residual heat removal service water (RHRSW) system
pump motor cooling solenoid valve; and
A trend of unavailability associated with the diesel-driven fire pump.
The inspectors reviewed applicable system maintenance rule scoping documents,
system health reports, corrective actions taken in response to the equipment problems,
maintenance rule functional failure determinations, and applicable a(1) action plans. In
addition, the issues were discussed with the responsible engineer.
b.
Findings
No findings of significance were identified.
1R13
Maintenance Risk Assessment and Emergent Work Evaluation (71111.13)
a.
Inspection Scope (seven samples)
The inspectors evaluated on-line and outage risk management for six planned and one
emergent maintenance activities. The inspectors reviewed maintenance risk
evaluations, work schedules, recent corrective actions, and control room logs to verify
that other concurrent or emergent maintenance activities did not significantly increase
plant risk. The inspectors also compared these items and activities to requirements
listed in Vermont Yankee AP 0125, "Equipment Release"; AP 0172, "Work Schedule
Risk Management - Online"; and AP 0173, Work Schedule Risk Management -
Outage. The inspectors reviewed the following work activities:
Online Risk:
Planned maintenance on the service water (SW) system supply to turbine the
building valve SW-19B breaker, resulting in Yellow online risk;
Planned maintenance on the A train of SBGT; and
Emergent work to implement minor modification on average power range
monitors (APRMs), resulting in a 1/2 scram condition and Yellow online risk.
7
Enclosure
Outage Risk:
Planned realignment and testing of offsite electrical power via the delayed
backfeed through the auxiliary and main transformers;
Planned maintenance resulting in 345 KV 340 line and 1T breaker being out of
service;
Portions of planned maintenance on electrical buses 2, 4, and 9; and
Planned performance of reactor pressure vessel leakage testing; considered by
Entergy to be a high risk evolution.
b.
Findings
No findings of significance were identified.
1R14
Personnel Performance During Non-routine Plant Evolutions (71111.14)
a.
Inspection Scope (two samples)
The inspectors assessed the control room operator performance during the following
two non-routine evolutions:
Entry into emergency operating procedure (EOP) 3, Primary Containment
Control, due to average torus temperature exceeding 90 degrees during HPCI
system testing on May 26, 2004; and
Reactor scram following the main transformer fire on June 18, 2004.
Specifically, the adequacy of personnel performance, procedure compliance, and use of
the corrective action process were evaluated against the requirements and expectations
contained in technical specifications and the following station procedures, as applicable:
AP 0151, Responsibilities and Authorities of Operations Department Personnel;
AP 0153, Operations Department Communication and Log Maintenance;
Vermont Yankee DP 0166, Operations Department Standards;
Vermont Yankee OP 105, Reactor Operations; and
OP 2124, Residual Heat Removal System.
b.
Findings
No findings of significance were identified.
1R15
Operability Evaluations (71111.15)
a.
Inspection Scope (five samples)
8
Enclosure
The inspectors reviewed five operability determinations prepared by the licensee. The
inspectors evaluated the selected operability determinations against the requirements
and guidance contained in NRC Generic Letter 91-18, Resolution of Degraded and
Nonconforming Conditions, as well as procedures AP 0167, Operability
Determinations, and ENN-OP-104, Operability Determinations. The inspectors
verified the adequacy of the following evaluations of degraded or non-conforming
conditions:
Flow noise from the C RHR system pump discharge orifice;
Broken 4 KV breaker driving pawl;
Missing clam shell from the control rod drive housing support system;
Apparent non-conservative flow-biased scram setpoints; and
Incomplete NDE for lifting and handling gear.
b.
Findings
No findings of significance were identified.
1R16
Operator Workarounds (71111.16)
a.
Inspection Scope (one sample)
The inspectors reviewed the cumulative effect of operator workarounds on the reliability,
availability, and potential mis-operation of systems and the potential to affect the ability
of operators to respond to plant transients and events. The inspectors reviewed
identified operator burdens, control room deficiencies, disabled or illuminated control
room alarms, and component deviations and discussed them with responsible
operations personnel to ensure they were appropriately categorized and tracked for
resolution. In addition, in-plant and control room tours were performed to identify any
workarounds not previously identified in accordance with procure DP 0166, Operations
Department Standards.
b.
Findings
No findings of significance were identified.
9
Enclosure
1R17
Permanent Plant Modifications
1.
Annual Review (71111.17A)
a.
Inspection Scope (one sample)
The inspectors performed an annual review of a permanent plant modification involving
the installation of an additional main steam safety valve installed during RFO 24. The
inspectors reviewed this modification to verify that the design bases, licensing bases,
and performance capability of risk significant structures, systems, and components
(SSCs) had not been degraded through the modifications. The review evaluated the
impact of the modification on power operation at the current licensed power level and
potential future operation at an increased power rating. This plant modification was
selected for review based on risk insights for the plant and included SSCs associated
with the initiating events, mitigating systems and barrier integrity cornerstones. The
inspection included a walkdown of the modification, interviews with plant staff, and the
review of applicable documents including procedures, Vermont Yankee Design
Calculation (VYDC) 2003-013, the modification package, engineering evaluations,
drawings, corrective action documents, the UFSAR and Technical Specifications. The
inspectors verified that selected attributes were consistent with the current design and
licensing bases. These attributes included component safety classification, energy
requirements supplied by supporting systems, instrument set-points, and control system
interfaces. Design assumptions were reviewed to verify that they were technically
appropriate and consistent with the UFSAR. The inspectors verified that selected
procedures, calculations and the UFSAR were properly updated with revised design
information and operating guidance. The inspectors also verified that the as-built
configuration was accurately reflected in the design documentation and that post-
modification testing was appropriate.
b.
Findings
No findings of significance were identified.
2.
Biennial Review (71111.17B)
a.
Inspection Scope (six samples)
The inspectors performed a biennial review of selected plant modifications that were
being installed during RFO 24. The modifications support a proposed power uprate that
is currently under review by the Office of Nuclear Reactor Regulation (NRR). The
inspectors reviewed the modifications to verify that the design bases, licensing bases,
and performance capability of risk significant SSCs had not been degraded through the
modifications. The reviews evaluated the impact of the modifications on power
operation at the current licensed power level and potential future operation at an
increased power rating. Plant modifications were selected for review based on risk
insights for the plant and included SSCs associated with the initiating events, mitigating
10
Enclosure
systems and barrier integrity cornerstones. The inspection included walkdowns of
selected plant systems and components, interviews with plant staff, and the review of
applicable documents including procedures, calculations, modification packages,
engineering evaluations, drawings, corrective action documents, the UFSAR and
Technical Specifications. The inspectors verified that selected attributes were
consistent with the current design and licensing bases. These attributes included
component safety classification, energy requirements supplied by supporting systems,
instrument set-points, and control system interfaces. Design assumptions were
reviewed to verify that they were technically appropriate and consistent with the UFSAR.
The inspectors verified that selected procedures, calculations and the UFSAR were
properly updated with revised design information and operating guidance. The
inspectors also verified that the as-built configuration was accurately reflected in the
design documentation and that post-modification testing was appropriate. A listing of
documents reviewed is provided in the Attachment to this report.
b.
Findings
No findings of significance were identified.
1R19
Post Maintenance Testing (71111.19)
a.
Inspection Scope (three samples)
The inspectors reviewed completed documentation for three post-maintenance test
(PMT) activities to verify the test data met the required acceptance criteria contained in
the licensees Technical Specifications, UFSAR, and in-service testing program, and
that the PMT was adequate to verify system operability and functional capability
following maintenance. The inspectors reviewed the PMTs performed after the following
maintenance activities:
Installation of low feedwater pump suction pressure trip modifications in
accordance with minor modification (MM) 2003-015;
APRM flow control trip reference card replacement in accordance with MM 2003-
028; and
Disassembly and repair of HPCI turbine exhaust check valve V23-3 following
failed as-found LLRT.
The inspectors verified that systems were properly restored following testing and that
discrepancies were appropriately documented in the corrective action process. The
inspectors also discussed the PMT results with the responsible engineers.
b.
Findings
No findings of significance were identified.
1R20
Refueling and Outage Activities (71111.20)
11
Enclosure
1.
Refueling Outage (RFO) 24
a.
Inspection Scope (one sample)
The inspectors evaluated the following outage activities to verify that Entergy considered
risk when developing outage schedules; that Entergy adhered to administrative risk
reduction methodologies for plant configuration control; and to ensure that Entergy
adhered to their operating license, Technical Specification requirements, and approved
procedures:
Review of the Outage Plan - The inspectors reviewed the RFO 24 shutdown risk
assessment to verify that Entergy addressed the outages impact on
defense-in-depth for the five shutdown critical safety functions; electrical power
availability, inventory control, decay heat removal, reactivity control, and
containment. Adequate defense-in-depth was verified for each safety function
and / or where redundancy was limited or not available, the existence of
appropriate planned contingencies, to minimize the overall risk, was verified.
Consideration of operational experience was also verified. The daily risk
up-date, accounting for schedule changes and unplanned activities were also
periodically reviewed;
Monitoring of Shutdown Activities - The inspectors observed the shutdown of the
reactor plant including reactor plant cooldown and transition to shutdown cooling
operations. As soon as practical following the shutdown, the inspectors
performed walkdowns of the primary containment;
Electrical Power - The inspectors reviewed the status and configuration of
safety-related buses throughout RFO 24. The inspectors ensured the electrical
lineups met the requirements of Technical Specification and the outage risk
control plan. The inspectors performed frequent walkdowns of affected portions
of the electrical plant including startup transformers, the auxiliary transformer,
and the emergency diesel generators;
Decay heat removal (DHR) System Monitoring - The inspectors monitored decay
heat removal status on a daily basis. Monitoring included daily reviews of
residual heat removal system alignment, reviews of spent fuel pool cooling
system alignment, and reviews of reactor coolant system (RCS) time-to-boil
calculations and results;
Inventory Control - The inspectors performed daily RCS inventory control reviews
including reviews of available injection systems and flow paths to ensure
consistency with the outage risk plan. The inspectors also ensured that
operators maintained reactor vessel and/or refueling cavity levels within
established ranges;
Reactivity Control - The inspectors observed reactivity management actions
taken by control room operators during refueling evolutions including procedure
place keeping, communications with refueling floor personnel, the monitoring of
source range nuclear instrumentation, and the monitoring of individual control
rod positions;
12
Enclosure
Containment Closure - The inspectors performed a torus internal cleanliness
walkdown following completion of outage activities. The inspectors performed a
primary containment closeout walkdown prior to final containment closure.
Finally, the inspectors ensured secondary containment was maintained as
required by Technical Specifications;
Refueling Activities - The inspectors observed portions of refueling operations,
including fuel handling and accounting in the reactor vessel and spent fuel pool.
The inspectors also performed an independent core reload verification of
approximately 34% of the core; and
Heatup and Startup Activities - The inspectors observed portions of the heatup
and startup of the reactor plant following the completion of RFO24.
The inspectors also verified that Entergy identified problems related to refueling
activities and entered them into their corrective actions program.
b.
Findings
Introduction: A very low safety significance (Green), self-revealing, non-cited violation
(NCV) of 10 CFR 50 Criterion XVI was identified in that Entergy personnel did not
develop effective corrective actions to prevent recurrence following a 2001 event
wherein control room operators did not verify a suction path existed prior to starting a
residual heat removal (RHR) system pump being used to support shutdown cooling
(SDC) operations which caused the pump to trip. On April 10, 2004, an identical event
occurred and again resulted in a trip of the RHR pump being used to support SDC
operations.
Description: On April 10, 2004, control room operators realigned vital alternating current
(AC) power from its normal power supply to the backup power supply to support planned
maintenance on a vital AC motor generator. The reactor plant was in the refueling
mode of operation at that time. In preparation for the vital AC realignment, operators
temporarily secured the RHR system, which was running in the SDC mode of operation.
One of the automatic actions that occurred during the vital AC alignment was the
closure of the RHR pump suction valve V10-17 from a Group 4 containment isolation
signal. Once the realignment of the vital AC power was completed, operators reset the
expected partial Group 4 containment isolation signal, but did not recognize that this
partial Group 4 containment isolation signal resulted in the closure of RHR system valve
V10-17, isolating the suction path used for RHR system support of SDC. Operators
subsequently attempted to reinitiate the RHR system in accordance with Vermont
Yankee Operating Procedure (OP) 2124, Residual Heat Removal System, Section J,
Short Term Shutdown Cooling Shutdown and Startup. When the B RHR pump was
started, the pumps breaker immediately tripped open due to a designed electrical
interlock requiring valve V10-17 to be open to provide a suction path for the RHR
system. Operators investigated the cause of the pump breaker trip, identified that no
suction path existed since valve V10-17 had closed, re-opened valve V10-17, and
successfully restarted the B RHR pump within 15 minutes of the breaker trip.
13
Enclosure
SDC thermal margin was maintained throughout this event via continued operation of
the spent fuel pool cooling system along with a calculated RCS time-to-boil value of
greater than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
In the apparent cause report for this event, Entergy identified that a nearly identical
event had occurred during a refueling outage in May 2001. At that time, operators had
performed a planned realignment of the vital AC power but did not recognize that valve
V10-17 had closed which resulted in a trip of the C RHR pump breaker when operators
attempted to reinitiate the RHR system. Entergy documented this previous event in
event report (ER) 2001-01228. Corrective actions assigned at that time included
discussions at shift supervisor meetings and the counseling of involved operators. In
the apparent cause report, Entergy also concluded that the corrective actions taken to
address the May 2001 event were insufficient to have prevented recurrence of the
nearly identical April 2004 event. Specifically, no corrective actions were assigned to
address the fact that OP 2124, Section J, did not specifically require operators to verify
an adequate RHR system flow path to and from the reactor existed prior to reinitiating
system operation.
Analysis: The performance deficiency associated with this finding is that Entergy
personnel did not assign effective corrective actions to prevent recurrence as required
by VY Administrative Procedure 0009 following a May 2001 trip of the C RHR pump
which occurred when operations did not recognize that RHR system valve V10-17 had
gone closed during a realignment of vital AC power. As a result, a similar event
occurred in April of 2004 involving a trip of the B RHR pump resulting from operators
again failing to recognize the closure of valve V10-17 during a realignment of vital AC
power. The finding is greater than minor since it is associated with the Fuel Cladding
Configuration Control Attribute of the Barrier Integrity Cornerstone and because it
affects the associated Cornerstone objective. Specifically, the April 2004 trip of the B
RHR pump, used to support SDC operations, reduced the assurance that the fuel
cladding would protect the public from radio nuclide releases caused by accidents or
events. The inspectors conducted a SDP Phase 1 screening of the finding in
accordance with IMC 0609, Appendix G, Shutdown Operations Significance
Determination Process [SDP]. The inspectors determined that Entergy did not meet
Item I.C. of Table 1, BWR [Boiling Water Reactor] Refueling Operation with RCS Level
> 23' since the finding resulted in Entergy not having at least one RHR loop operating
to support SDC. However, the inspectors also determined that the finding did not
degrade Entergys ability to recover SDC since the B RHR pump was restarted within
15 minutes of being tripped and an adequate thermal margin was maintained via a
calculated RCS time-to-boil of greater than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. Therefore, in accordance with
IMC 0609, Appendix G, the finding was of very low safety significance (Green).
A contributing cause of this finding is related to the Cross-Cutting area of Problem
Identification and Resolution. As stated above, Entergy personnel did not develop
effective corrective actions to prevent recurrence following a 2001 event wherein control
room operators did not verify a suction path existed prior to starting the RHR system
pump being used to support SDC operations which caused the pump to trip. Entergys
corrective actions relied on the operators skill to verify a suction path was open prior to
14
Enclosure
restarting the RHR pump rather than proceduralize the step. As a result, an identical
event occurred in April 2004 again resulting in a trip of the RHR pump being used to
support SDC operations.
Enforcement:
10 CFR 50, Appendix B, Criterion XVI states, in part, that measures shall be established
to assure that conditions adverse to quality are promptly identified and corrected.
Vermont Yankee AP 0009, Event Reports, Revision 12, describes Entergys
requirements for the identification and correction of conditions adverse to quality
including determining the cause(s) of the event and assigning corrective actions that
prevent recurrence. Contrary to the above, in May 2001, Entergy did not assign
effective corrective actions that prevent recurrence following a May 2001 trip of the C
RHR pump which occurred when operators did not recognize that RHR system valve
V10-17 had closed due to an expected partial Group 4 containment isolation during the
realignment of vital AC power. As a result, a similar event occurred in April of 2004
involving the trip of the B RHR pump resulting from operators again failing to recognize
the closure of valve V10-17 during a realignment of vital AC power. Because the finding
is of very low safety significance and has been entered into the licensees Corrective
Actions Program (CR 2004-01005), this violation is being treated as an NCV, consistent
with Section VI.A of the NRC Enforcement Policy: NCV 0500271/2004003-01,
Ineffective Corrective Actions Assigned Following a May 2001 Trip of the C RHR
System Pump During SDC Operation.
2.
Forced Outage Following the Main Transformer Fire of June 18, 2004.
a.
Inspection Scope (partial sample)
The inspectors evaluated the following forced outage activities to verify that Entergy
considered risk when developing outage schedules; that Entergy adhered to
administrative risk reduction methodologies for plant configuration control; and to ensure
that Entergy adhered to their operating license, Technical Specification requirements,
and approved procedures:
Review of the Outage Plan - The inspectors reviewed the shutdown risk
assessment to verify that Entergy addressed the outages impact on
defense-in-depth for the five shutdown critical safety functions; electrical power
availability, inventory control, decay heat removal, reactivity control, and
containment. The daily risk up-date, accounting for schedule changes and
unplanned activities were also periodically reviewed;
Monitoring of Shutdown Activities - The inspectors observed the shutdown of the
reactor plant including reactor plant cooldown activities and transition to
shutdown cooling operations. As soon as practical following the shutdown, the
inspectors performed walkdowns of the primary containment;
DHR System Monitoring - The inspectors monitored decay heat removal on a
daily basis. Monitoring included daily reviews of residual heat removal system
15
Enclosure
alignment, reviews of spent fuel pool cooling system alignment, and reviews of
RCS time-to-boil calculations and results; and
Inventory Control - The inspectors performed daily RCS inventory control reviews
including reviews of available injection systems and flow paths to ensure
consistency with the outage risk plan. The inspectors also ensured that
operators maintained RCS level within established ranges.
The inspectors also verified that Entergy identified problems related to the forced outage
and entered them into their corrective actions program.
b.
Findings
No findings of significance were identified.
1R22
Surveillance Testing (71111.22)
a.
Inspection Scope (eight samples)
The inspectors observed surveillance testing to verify that the test acceptance criteria
specified for each test was consistent with Technical Specification and UFSAR
requirements, was performed in accordance with the written procedure, the test data
was complete and met procedural requirements, and the system was properly returned
to service following testing. The inspectors observed selected pre-job briefs for the test
activities. The inspectors also verified that discrepancies were appropriately
documented in the corrective action program. The inspectors verified that testing in
accordance with the following procedures met the above requirements:
OP 4031, Type B and C Primary Containment Leak Rate Calculations and
Evaluations;
OP 4100, ECCS Integrated Automatic Initiation Test;
OP 4114, Standby Liquid Control [SLC] System Surveillance, Section C, Flow
Test Directly into the Reactor Vessel, and Section I, SLC Explosive Charge
Continuity Check;
OP 4121, Reactor Core Isolation Cooling System Surveillance, Section B,
RCIC Injection Check Valve (RCIC-22) Test;
OP 4142, Vernon Tie and Delayed Access Power Source Backfeed
Surveillance;
OP 4424, Control Rod Scram Testing and Data Reduction, Section B, Single
Rod Scrams Using ERFIS Data Collection;
OP 4430, Reactivity Anomalies/Shutdown Margin Check, Section 1, Strongest
Control Rod Withdrawn Subcritical Check; and
Special Test Procedure (STP) 2003-004, Power Ascension Test Procedure.
b.
Findings
No findings of significance were identified.
16
Enclosure
1R23
Temporary Modifications (71111.23)
a.
Inspection Scope (two samples)
The inspectors reviewed the following temporary modifications (TMs) to ensure that the
modifications did not adversely affect the availability, reliability, or functional capability of
any risk-significant structures, systems, and components:
TM 2003-039, Bottom Head Drain Line Freeze Seal; and
TM 2003-022, Vibration Monitoring Equipment Installation on MS & FW Piping.
The inspectors compared the information in the TM packages to Entergys TM
requirements contained in AP 0020, Control of Temporary and Minor Modifications.
The inspectors also walked down accessible portions of these TMs to verify that
required tags and markings were applied and that the TMs were properly maintained.
The inspectors also reviewed a sample of TM-related problems identified in the
Entergys corrective action program to verify that they had identified and implemented
appropriate corrective actions.
b.
Findings
No findings of significance were identified.
Cornerstone: Emergency Preparedness
1EP6
Drill Evaluation (71114.06)
a.
Inspection Scope (one sample)
On June 17, 2004, the inspectors observed an operating crew evaluate a simulator-
based event using the station emergency action levels (EALs) during licensed operator
requalification training activities. The inspectors discussed the performance
expectations and results with the lead instructor and operations training manager. The
inspectors focused on the ability of licensed operators to perform event classification
and make proper notifications in accordance with the following station procedures and
industry guidance:
AP 0153, Operations Department Communications and Log Maintenance;
AP 0156, Notification of Significant Events;
AP 3125, Emergency Plan Classification and Action Level Scheme;
DP 0093, Emergency Planning Data Management;
OP 3540, Control Room Actions During an Emergency; and
Nuclear Energy Institute (NEI) 99-02, Regulatory Assessment Performance
Indicator Guideline, Revision 2.
b.
Findings
17
Enclosure
No findings of significance were identified.
2.
RADIATION SAFETY
Cornerstone: Occupational Radiation Safety
2OS1 Access Control to Radiologically Significant Areas (71121.01)
a.
Scope (fourteen samples)
The inspectors conducted inspections to verify that Entergy was properly implementing
physical, engineering, and administrative controls for access to high radiation areas, and
other radiologically controlled areas, and that workers were adhering to these controls
when working in these areas. Implementation of the access control program was
reviewed against the criteria contained in 10 CFR 20, Technical Specifications, and
approved Entergy procedures. The inspectors conducted independent radiation surveys
and observed work area conditions, reviewed radiation surveys of these areas, and
reviewed electronic dosimetry set points and other exposure controls specified in the
radiation work permits (RWPs) that provided the access control requirements for the
following radiologically significant work activities:
Steam dryer underwater welding modifications;
Drywell shielding installation;
Drywell in-service inspection of core spray nozzle N5A;
Drywell safety relief valve maintenance;
Drywell main steam isolation valve maintenance; and
Feedwater heater replacement modifications
b.
Findings
No findings of significance were identified.
2OS2 ALARA Planning and Controls (71121.02)
Inspection Scope (four samples)
The inspectors reviewed Entergys As Low As Reasonably Achievable (ALARA)
Program performance against the requirements of 10 CFR 20.1101(b). The inspectors
reviewed aspects of the implementation of exposure reduction requirements based on
ALARA planning for the five highest exposure outage tasks. The ALARA-related work
activities observed are listed in Section 2OS1 above. In addition, the following ALARA
inspection activities were conducted:
Independent shielding effectiveness radiation surveys conducted in the drywell;
Observation of closed circuit television equipment and tele-dosimetry use in the
drywell was conducted with respect to drywell remote health physics work
surveillance capability and technical specification requirements; and
18
Enclosure
Feedwater heater bay source term location was reviewed relative to worker
occupancy areas.
b.
Findings
No findings of significance were identified.
4.
OTHER ACTIVITIES (OA)
4OA1 Performance Indicator Verification (71151)
a.
Inspection Scope (two samples)
The inspectors sampled Entergy submittals for the performance indicators (PIs) listed
below for the period from April 2003 to March 2004. The PI definitions and guidance
contained in NEI 99-02 and AP 0094, NRC Performance Indicator Reporting, were
used to verify the accuracy and completeness of the PI data reported during this period.
Barrier Integrity Cornerstone
Reactor Coolant System Specific Activity; and
Reactor Coolant System Leakage.
The inspectors reviewed selected operator logs, plant process computer data, condition
reports, and monthly operating reports for the period April 1, 2003, through March 31,
2004.
b.
Findings
No findings of significance were identified.
4OA2 Identification and Resolution of Problems (71152)
1.
Routine Review of Identification and Resolution of Problems
a.
Inspection Scope
The inspectors routinely reviewed issues during baseline inspection activities and plant
status reviews to verify they were being entered into Entergys corrective action system
at an appropriate threshold and that adequate attention was being given to timely
corrective actions. Additionally, in order to identify repetitive equipment failures and/or
specific human performance issues for follow-up, the inspectors performed a daily
screening of items entered into Entergys corrective action program. This review was
accomplished by reviewing selected hard copies of condition reports (a listing of CRs
reviewed is included in the Attachment to this report) and/or by attending daily screening
meetings.
19
Enclosure
b.
Findings
No findings of significance were identified.
2.
Semi-Annual Trend Review
a.
Inspection Scope
As required by Inspection Procedure 71152, Identification and Resolution of Problems,
the inspectors performed the semi-annual trend review to identify trends, either licensee
or NRC identified, that might indicate the existence of a more significant safety issue.
Included within the scope of this review were:
CRs generated from January through May 2004;
Corrective maintenance backlog listings from January through May 2004;
The corrective action program 3rd and 4th quarter, 2003 trend report; and
Daily review of main control room operator logs.
b.
Findings
No findings of significance were identified.
3.
Cross-Reference to PI&R Findings Documented Elsewhere
Section 1R20.1 describes a finding wherein Entergy personnel did not develop effective
corrective actions to prevent recurrence following a 2001 event wherein control room
operators did not verify a suction path existed prior to starting the RHR system pump
being used to support SDC operations which caused the pump to trip. Entergys
corrective actions relied on the operators skill to verify a suction path was open prior to
restarting the RHR pump rather than proceduralize the step. As a result, an identical
event occurred in April 2004 again resulting in a trip of the RHR pump being used to
support SDC operations.
4OA3 Event Followup (71153)
1.
Main Transformer Fire and Reactor Plant Scram
a.
Inspection Scope (1 sample)
The inspectors evaluated Entergys response to a main transformer fire and resultant
reactor plant scram that occurred on June 18, 2004. The inspectors immediately
responded to the main control room to observe reactor plant parameters, to evaluate
individual safety system responses, and to evaluate licensed operator responses to the
event. The inspectors evaluated the response of the reactor plant and the licensed
operators against Entergy approved operating procedures, abnormal operating
procedures, and emergency operating procedures. The inspectors evaluated Entergys
classification of the event (i.e., Unusual Event) in accordance with approved EAL
20
Enclosure
procedures to ensure notifications were made to NRC and state/county governments as
required. The inspectors also evaluated the ability of Entergys fire brigade and
automatic fire protection systems to extinguish the main transformer fire in a safe and
timely manner.
The NRC Region I Office dispatched two inspectors, each a specialist in the areas of
electrical and fire protection systems, to assist the resident inspectors with event follow-
up activities. The inspectors monitored Entergys efforts in determining the root cause
of the event; monitored Entergys efforts for the recovery, replacement, and repair of the
effected portions of the 22KV electrical system; and monitored Entergys reactor plant
restart preparation activities.
b.
Findings
Entergy has identified that the root cause of the main transformer fire relates to
weaknesses with the preventive maintenance performed on the 22 KV electrical system.
Because additional information is needed to determine if these issues are more than
minor, they are considered to be an unresolved item (URI) pending completion of the
inspectors review of Entergys root cause analysis: URI 0500271/2004003-02,
Weaknesses Identified with the Preventive Maintenance Performed on the 22 KV
Electrical System Resulted in Main Transformer Fire.
4OA5 Other Activities
1.
Temporary Instruction (TI) 2515/156, Offsite Power System Operational Readiness.
a.
Inspection Scope
The inspectors collected and reviewed information pertaining to the Vermont Yankee
offsite power system as it related to the requirements of 10 CFR 50.65, Requirements
for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants; 10 CFR 50.63, Loss of All Alternating Current Power; offsite power operability; and corrective
actions. The inspectors also reviewed this data against the requirements of 10 CFR 50,
Appendix A, General Design Criterion 17, Electric Power Systems, and the Vermont
Yankee Technical Specifications. This information was forwarded to NRR for further
review. A listing of documents reviewed is included in the Attachment to this report.
b.
Findings
No findings of significance were identified.
4OA6 Meetings, including Exit
Resident Exit
On July 12, 2004, the resident inspectors presented the inspection results to Mr. Kevin
Bronson and members of his staff. The inspectors asked whether any materials
21
Enclosure
examined during the inspection should be considered proprietary. No proprietary
information was identified.
Meeting with the State of Vermont Public Service Board
On June 28, 2004, Region I and NRR staff met with the Vermont State Public Service
Board (PSB) regarding Vermont Yankees request for a 20% extended power uprate.
The NRC staff discussed the NRCs power uprate review process and details regarding
a planned pilot engineering inspection slated for Vermont Yankee in August 2004.
ATTACHMENT: SUPPLEMENTAL INFORMATION
A-1
Attachment
SUPPLEMENTAL INFORMATION
KEY POINTS OF CONTACT
Licensee Personnel:
J. Thayer
Site Vice President
K. Bronson
General Plant Manager
J. Allen
Design Engineering
P. Corbett
Maintenance Manager
J. Dreyfuss
Project Engineering Manager
J. Devincentis Licensing Manager
W. Fadden
Design Engineering
J. Geyster
Radiation Protection Superintendent
D. Giorowall
Programs Supervisor
Dennis Girrior Programs Supervisor
S. Goodwin
Mechanical Design Department Manager
M. Gosekamp Superintendent of Operations Training
M. Hamer
Licensing
D. Johnson
Design Engineering
Dave King
ISI Coordinator
R. Morissette Principal As Low As Reasonably Achievable (ALARA) Engineer
M. Pletcher
Radiation Protection Supervisor - Instruments
P. Rainey,
Design Engineering
B. Renny
Supervisor, Access Authorization
K. Stupak
Technical Training
C. Wamser
Operations Manager
R. Wanczyk
Director of Nuclear Safety
LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED
Opened and Closed
Ineffective Corrective Actions Assigned Following a May
2001 Trip of the C RHR System Pump During SDC
Operation (Section 1R20.1)
Opened
Weaknesses Identified with the Preventive Maintenance
Performed on the 22 KV Electrical System Resulted in
Main Transformer Fire (Section 4OA3.1)
A-2
Attachment
LIST OF DOCUMENTS REVIEWED
Section 1R02: Evaluation of Changes, Tests, or Experiments
Power Uprate Modifications
TM 2003-022
Vibration Monitoring Equipment Installation on MS [Main Steam] & FW
[Feedwater] Piping
MM 2003-015
Reactor Feed Pump Suction Pressure Trip
MM 2003-016
Reactor Recirculation System Runback
MM 2003-026
AST [Alternate Source Term] Component Modification (OG-779
Installation)
MM 2003-028
APRM Flow Control Trip Reference Card Replacement
MM 2003-039
NSSS [Nuclear Steam Supply System]/BOP [Balance of Plant]
Instrumentation Upgrades
MM 2003-054
381 Line Overload Relay Setting
VYDC 2003-013
Installation of Additional Main Steam Safety Valve
Section 1R08: Inservice Inspection
Procedures
ENN-NDE 9.29, Rev 0 for UT of structural overlay (weld N5A)
PDI-UT-8, Rev B. Generic Procedure for UT of Weld Overlaid Austenitic Pipe Welds
ISI - 254, Rev 5, for remote ISI of RPV Welds
NE 8048, Rev 1 - In Vessel Visual Inspection
Drawings
ISI-PPV-103, Rev 3. Reactor Vessel
ISI-SLC-Part 4, Rev 3. SLC Piping ISO
D-7983-621 Rev G. UT/ET clad crack calibration block
6D30047, Rev 0, Wesdyne Calibration Standard PDI-01
Miscellaneous Reports
QA (Quality Assurance) Audit Report AR-2003-22b&c, dated 11/13/2003
GE (General Electric) RICSIL No. 050 of 4/23/1990, and GE SIL NO. 539, dated 11/5/1991
GE Reports INR-VYR24-04-01R2, 02R2, 03, & 04R1 on Steam Dryer Visual Indications
GE Nuclear Engineering (GENE) 0000-0028-0130-01, Revision 3, dated April 2004 on Steam
Dryer Unit End Plate Indications - Vermont Yankee R24
GENE-0000-0028-0130-02, Revision 3, dated April 2004 on Steam Dryer Drain Channel
Indications - Vermont Yankee R24
Section 1R17: Permanent Plant Modifications
Power Uprate Modifications
A-3
Attachment
MM 2003-015
Reactor Feed Pump Suction Pressure Trip
MM 2003-016
Reactor Recirculation System Runback
MM 2003-026
AST Component Modification (OG-779 Installation)
MM 2003-028
APRM Flow Control Trip Reference Card Replacement
MM 2003-039
NSSS/BOP Instrumentation Upgrades
MM 2003-054
381 Line Overload Relay Setting
Calculations
Vermont Yankee Calculation (VYC) 0693A Rev. 2 APRM Neutron Monitoring Trip Loops
VYC-2269 Rev. 0
Feedwater and Condensate Hydraulic Model Analysis
VYC-2309 Rev. 0
Steam Drain Line MS-189-D3 Check Valve Addition
License Amendment Documents
License Amendment Proposal for ARTS/MELLLA
Technical Specification Proposed Change # 257 (ARTS/MELLLA)
GE-NE-0000-0020
Entergy Nuclear Operations Incorporated Vermont Yankee Nuclear
Power
GE-NE-1500-0001
Station MELLLA+ Transient Analysis
Safety Analysis Report for Vermont Yankee Nuclear Power Station
Constant Pressure Power Uprate
NRC NRR Safety Evaluation for License Amendment No. 219 to DPR-28
Specifications/Procedures
AP 5226 Rev. 5
Calibration of Switchyard Breaker Failure Relays
VYSP-FS-074
Specifications for Safety Valves
Individual Plant Examination for SRV/SV Reclosure
Section 4OA2.1: Routine Review of Problem Identification and Resolution
Condition Reports
2002-2581
RBCCW pumps failed to restart within time limit during ECCS [emergency core
isolation cooling] test
2002-2584
ECCS test data was accepted as satisfactory when some data was outside of
acceptance criteria
2003-1509
The C RHRSW pump cooling water supply solenoid valve failed to open as
required on pump start
2003-2321
No indicated cooling flow upon C RHRSW pump start
2004-0700
While troubleshooting a 4KV breaker on Bus-2-7, the breaker driving pawl broke
- 2004-0840
Incorrect status of Decay Heat Removal was logged on the Critical Outage
Systems Status Form
- 2004-0845
NRC resident question on RHR procedure wording
2004-0879
HPCI V23-845 failed IST testing
2004-0892
Water level in the reactor cavity exceeds limits during cavity floodup
A-4
Attachment
- 2004-0897
Incorrect start dates used in ORAM-Sentinel for alternate DHR capability
determinations
2004-0918
Adverse trend - main steam isolation valve Appendix J test failures
2004-0942
HPCI V23-846 failed IST testing
2004-0955
As-found condition of V2-80 included a galled stem
2004-0968
Unsuccessful decon of diver
2004-0981
An observation was made from below vessel that a piece of control rod drive
housing support (shoot-out steel) was missing
2004-0986
Instructions for RWP not adhered to
2004-0998
RHR-46A allowed to overflow while working on the valve
2004-1005
B RHR pump trip during restart due to no suction path
2004-1017
V2-13-3 failed Appendix J local leak rate test
2004-1058
Flow noise from RO-10-105C, C RHR pump discharge orifice
2004-1091
Rad survey maps indicate need to perform alpha survey
2004-1117
Flow noise from C RHR pump discharge orifice
2004-1160
ASME rejectable indication on SLC weld
2004-1190
Weld electrodeoven left unlocked and unattended
- 2004-1339
Two fuel segments could not be confirmed in storage container
2004-1409
A RBCCW did not start within the allowed ECCS start time
2004-1426
ECCS test exceptions
2004-1428
Reactor water clean up pump started with no suction path
2004-1548
P-8-1A leaking oil from upper bearing reservoir area
2004-1653
Excessive overtime approved without documentation
2004-1665
Potentially non-conservative scram setpoint values
- 2004-1916
- 2 fan room has inadequate hose stream coverage due to modification to fan
room door
- 2004-1928
Slight leakage on B SBGT demister loop seal piping union
2004-1989
Generator Ground alarm came in
2004-2015
Reactor Scram
2004-2017
Notification of Unusual Event (NOUE) declared due to plant fire and automatic reactor scram
2004-2019
Main transformer fire
- 2004-2022
Discrepancy in post scram rod position indication
- 2004-2023
Torus-to-drywell vacuum breaker indicating lights and alarm indicate breakers
may have cycled during the scram/transformer trip
- 2004-2045
Repeat of P-8-1A leaking oil from upper bearing reservoir area
2004-2074
Failure to make timely notification of States upon declaration of unusual event on
June 18, 2004
- Inspector-identified issues.
Section 4OA5.1: Temporary Instruction (TI) 2515/156, Offsite Power System Operational
Readiness.
Procedures
A-5
Attachment
Vermont Yankee Operating Procedure Form (VYOPF) 0150.03, CRO [Control Room Operator]
Round Sheet
AP 0172, Work Schedule Risk Management - On Line
ISO New England Master/Satellite Procedure #1, Nuclear Plant Transmission Operations,
Revision 0
ISO New England Master/Satellite Procedure #2, Abnormal Conditions Alert, Revision Dated
11/19/01
Licensee Event Reports (LERs)
Vermont Yankee Nuclear Power Station LER 87-008-00, Loss of Normal Power During
Shutdown Due to Routing All Off-Site Power Sources Through One Breaker
Vermont Yankee Nuclear Power Station LER 84-022-00, Diesel Generator Lockout Trip of
Both Generators
Maintenance Rule Documents
NRC Maintenance Rule Program Website Frequently Asked Questions (FAQs)
Vermont Yankee 10CFR50.65 NRC Maintenance Rule SSC Basis Document, 345K Volts AC
Electrical (345KV)
Vermont Yankee 10CFR50.65 NRC Maintenance Rule SSC Basis Document, 115K Volts AC
Electrical (115KV)
Operational Experience Documents
JA Fitzpatrick Operational Experience (OE) 16822, Reactor Scram due to Grid Instability
Significant Operating Experience Report (SOER) 9901, Loss of Grid
Miscellaneous Documents
Control room operator logs dated 8/17/87
VYC-1088, Vermont Yankee 4160/480 Volt Short Circuit/Voltage Study, Revision 3
A-6
Attachment
LIST OF ACRONYMS
Alternating Current
Automated Document Access Management System
As Low As Is Reasonably Achievable
Vermont Yankee Administrative Procedure
Average Power Range Monitors
American Society of Mechanical Engineers
CFR
Code of Federal Regulations
CR
Condition Report
CRO
Control Room Operator
CY
Calendar Year
Design Basis Document
DOST
Diesel Oil Storage Tank
DP
Vermont Yankee Department Procedure
Emergency Action Levels
Emergency Operating Procedure
ER
Event Report
Frequently Asked Question
Main Feedwater System
GENE
General Electric Nuclear Engineering
High Pressure Coolant Injection
IMC
Inspection Manual Chapter
Individual Plant Examination External Events
IR
Inspection Report
Inservice Inspection
Inservice Testing
KV
Kilovolt
LER
Licensee Event Report
Local Leakage Rate Testing
MM
Minor Modification
MS
Main Steam System
Non-Cited Violation
NEI
Nuclear Engineering Institute
Notice of Unusual Event
NRC
Nuclear Regulatory Commission
NRC Office of Nuclear Reactor Regulation
Operating Experience
ON
Vermont Yankee Off-Normal Procedure
OP
Vermont Yankee Operating Procedure
A-7
Attachment
Performance Indicator
Post Maintenance Testing
PSB
Public Service Board
Quality Assurance
Reactor Core Isolation Cooling
Refueling Outage
Residual Heat Removal Service Water
Radiation Work Permit
Standby Gas Treatment
Significance Determination Process
SEG
Simulator Evaluation Guide
SEN
Significant Event Notification
Spent Fuel Pool
Significant Operating Experience Report
Structures, Systems and Components
Special Test Procedure
TI
Temporary Instruction
TM
Updated Final Safety Analysis Report
Unresolved Item
Ultrasonic Testing
Visual Examination Testing
Vermont Yankee
VYC
Vermont Yankee Calculation
VYDC
Vermont Yankee Design Calculation
VYOPF
Vermont Yankee Operating Procedure Form