ML110280456
ML110280456 | |
Person / Time | |
---|---|
Site: | Watts Bar |
Issue date: | 01/28/2011 |
From: | Eugene Guthrie Reactor Projects Region 2 Branch 6 |
To: | Krich R Tennessee Valley Authority |
References | |
IR-10-005 | |
Download: ML110280456 (34) | |
See also: IR 05000390/2010005
Text
UNITED STATES
NUCLEAR REGULATORY COMMISSION
REGION II
245 PEACHTREE CENTER AVENUE NE, SUITE 1200
ATLANTA, GEORGIA 30303-1257
January 28, 2011
Mr. R. M. Krich
Vice President, Nuclear Licensing
Tennessee Valley Authority
3R Lookout Place
1101 Market Street
Chattanooga, TN 37402-2801
SUBJECT: WATTS BAR NUCLEAR PLANT - NRC INTEGRATED INSPECTION REPORT
Dear Mr. Krich:
On December 31, 2010, the United States Nuclear Regulatory Commission (NRC) completed
an inspection at your Watts Bar Nuclear Plant, Unit 1. The enclosed integrated inspection
report documents the inspection results which were discussed on January 10, 2010, with Mr. D.
Grissette 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 three NRC-identified findings which were determined to be of very low
safety significance (Green). These findings were determined to involve violations of NRC
requirements. However, because of their very low safety significance and because they are
entered into your corrective action program, the NRC is treating these findings as non-cited
violations (NCVs) consistent with the NRC Enforcement Policy. If you contest any NCV in this
report, you should provide a response within 30 days of the date of this inspection report, with
the basis for your denial, to the Nuclear Regulatory Commission, ATTN.: Document Control
Desk, Washington DC 20555-0001; with copies to the Regional Administrator, Region II; the
Director, Office of Enforcement, United States Nuclear Regulatory Commission, Washington,
DC 20555-0001; and the NRC Resident Inspector at the Watts Bar facility.
TVA 2
In accordance with 10 CFR 2.390 of the NRC's "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/
Eugene F. Guthrie, Chief
Reactor Projects Branch 6
Division of Reactor Projects
Docket Nos.: 50-390
License No.: NPF-90
Enclosure: NRC Inspection Report 05000390/2010005
w/Attachment: Supplemental Information
cc w/encl: (See page 3)
_ ML110280456__ G SUNSI REVIEW COMPLETE
OFFICE RII:DRP RII:DRP RII:DRS RII:DRP RII:DRS RII:DRS RII:DRS
SIGNATURE RLM /RA/ Via email BBD /RA for/ EFG /RA for/ MKM /RA for/ Via email BBD /RA for/
NAME RMonk WDeschaine PHiggins MSchwieg RBaldwin MMeeks RLewis
DATE 01/26/2011 01/26/2011 01/28/2011 01/28/2011 01/28/2011 01/27/2011 01/28/2011
E-MAIL COPY? YES NO YES NO YES NO YES NO YES NO YES NO YES NO
OFFICE RII:DRS RII:DRP RII:DRP
SIGNATURE BBD /RA for/ CRK /RA/ EFG /RA/
NAME RWilliams CKontz EGuthrie
DATE 01/28/2011 01/28/2011 01/28/2011
E-MAIL COPY? YES NO YES NO YES NO YES NO YES NO YES NO YES NO
TVA 3
cc w/encl:
D. E. Grissette
Vice President
Watts Bar Nuclear Plant
Tennessee Valley Authority
P.O. Box 2000
Spring City, TN 37381
G. A. Boerschig
Plant Manager
Watts Bar Nuclear Plant
Tennessee Valley Authority
P.O. Box 2000
Spring City, TN 37381
M. K. Brandon
Manager
Licensing and Industry Affairs
Watts Bar Nuclear Plant
Electronic Mail Distribution
E. J. Vigluicci
Assistant General Counsel
Tennessee Valley Authority
6A West Tower
400 West Summit Hill Drive
Knoxville, TN 37902
County Mayor
P.O. Box 156
Decatur, TN 37322
Ann Harris
341 Swing Loop
Rockwood, TN 37854
TVA 4
Letter to R. M. Krich from Eugene Guthrie dated January 28, 2011
SUBJECT: WATTS BAR NUCLEAR PLANT - NRC INTEGRATED INSPECTION REPORT
Distribution w/encl:
C. Evans, RII
L. Douglas, RII
OE Mail
RIDSNRRDIRS
PUBLIC
RidsNrrPMWattsBar1 Resource
RidsNrrPMWattsBar2 Resource
U.S. NUCLEAR REGULATORY COMMISSION
REGION II
Docket No: 50-390
License No: NPF-90
Report No: 05000390/2010005
Licensee: Tennessee Valley Authority (TVA)
Facility: Watts Bar Nuclear Plant, Unit 1
Location: Spring City, TN 37381
Dates: October 1 - December 31, 2010
Inspectors: R. Monk, Senior Resident Inspector
W. Deschaine, Regional Inspector, Region II (RII)
P. Higgins, Regional Inspector, RII
M. Schwieg, Resident Inspector
R. Baldwin, Senior Operations Engineer (1R11.2, 3)
M. Meeks, Operations Engineer (1R11.3)
R. Lewis, Resident Inspector (4OA5.2, 3)
R. Williams, Reactor Inspector (4OA5.1)
Approved by: Eugene F. Guthrie, Chief
Reactor Projects Branch 6
Division of Reactor Projects
Enclosure
SUMMARY OF FINDINGS
IR 05000390/2010-005; 10/01/2010 - 12/31/2010; Watts Bar, Unit 1; Maintenance Effectiveness
and Other Activities
The report covered a three-month period of routine inspection by resident inspectors. Three
NRC identified findings, each of which are non-cited violations (NCVs), were identified. The
significance of an issue is indicated by its color (Green, White, Yellow, Red) using the
Significance Determination Process in Inspection Manual Chapter 0609, Significance
Determination Process (SDP). The NRCs program for overseeing the safe operation of
commercial nuclear power reactors is described in NUREG-1649, Reactor Oversight Process,
Revision 4, dated December 2006.
A. NRC-Identified Findings and Self-Revealing Findings
Cornerstone: Mitigating Systems
- Green. The inspectors identified a non-cited violation of 10 CFR 50.65(a)(2), was
identified by the inspectors for the licensees failure to set goals and monitor the
performance and condition of the B Main Control Room (MCR) Air Conditioning
system as required by 10CFR50.65(a)(1), and had no justification for not doing so,
after it had failed to demonstrate effective control of the performance or condition of
the system through appropriate preventive maintenance. The inspectors identified
three Component Deficiency Reports that documented failures which had been
evaluated by the licensee as non-functional failures. The licensee has subsequently
implemented goal setting and monitoring requirements specified in 10 CFR
50.65(a)(1) and entered this issue into the corrective action program as PER
205438.
The inspectors determined that this finding was more than minor since the B MCR
Air Conditioning Train was not placed in (a)(1) monitoring status in a timely manner
which if left uncorrected, could become a more significant safety concern. NRC staff
review has determined this MR violation to have a very low safety significance
(Green) because it was not among the contributing causes of the degraded
performance and condition of the B Main Control Room (MCR) Air Conditioning
system and not processed through the significance determination process. The
cause of the finding was directly related to the cross-cutting area of Problem
Identification and Resolution, evaluation aspect of the corrective action program
component, in that, the licensee failed to thoroughly evaluate failures and determine
those failures to be functional failures of the B MCR Air Conditioning System such
that the system was placed in category a(1) in a timely manner. P.1(c) (Section
1R12)
- Green. The inspectors identified a non-cited violation of 10 CFR 50, Appendix B,
Criterion III, Design Control, for the failure to assure that appropriate quality
standards were specified and included in design documents and that deviations from
such standards were controlled. Specifically, the licensee failed to demonstrate the
necessary conditions for commercial grade dedication and seismic qualification of
Enclosure
3
molded case circuit breakers to safety-related application within the station 120VAC
vital instrumentation boards. Corrective actions for this issue are still being
evaluated and has been entered into the licensees corrective action program as
PER 171695.
Failure to specify appropriate qualification standards in performing commercial grade
dedication of a component-level commodity is a performance deficiency. This
performance deficiency is more than minor and a finding because it affected the
design control attribute of the mitigating systems cornerstone objective to ensure the
availability, reliability, and capability of systems that respond to initiating events to
prevent undesirable consequences. Specifically, adequate measures were not
implemented to ensure the station 120VAC vital instrumentation boards were
properly seismically qualified for their application. The inspector assessed the finding
using the SDP and determined that the finding was of very low safety significance
(Green) because the breaker panels had originally been qualified by testing a
complete prototype panel, while the licensees processes replaced a component-
level item within that panel utilizing the original make and model component through
commercial grade dedication. The inspectors concluded that overall operability was
not brought into question.
This finding was reviewed for cross-cutting aspects and none were identified, as it
was determined not to reflect current licensee performance. (Section 4OA5.2)
- Green. The inspectors identified a non-cited violation of 10 CFR 50, Appendix B,
Criterion III, Design Control, for the failure to assure that applicable regulatory
requirements and the design basis for structures, systems, and components are
correctly translated into specifications, drawings, procedures, and instructions.
Specifically, the licensee failed to assure that applicable regulatory requirements for
undervoltage (degraded) voltage protection, including those prescribed in TS 3.3.5-1,
item 2, were correctly translated into design calculation, WBN-EEB-MS-TI-06-0029,
Degraded Voltage Analysis, Revision. 31, which evaluated motor starting voltages
at the beginning of a design basis loss of coolant accident (LOCA) concurrent with a
degraded grid condition. Corrective actions for this issue are still being evaluated
and has been entered into the licensees corrective action program as PER 296306.
The failure to use the degraded voltage relay setpoint values as specified in TS and
configured in the 6900 VAC bus based on the electrical design calculation was a
performance deficiency. This finding is more than minor because it affects the
Design Control attribute of the Mitigating Systems Cornerstone. It impacts the
cornerstone objective of ensuring the availability, reliability, and operability of the
6900 VAC safety buses to perform the intended safety function during a design basis
event. The potential availability, reliability, and operability of the 6900 VAC safety
buses during a potential degraded voltage condition was impacted as the licensee
design calculation used a non-conservative degraded voltage input, with respect to
the values specified in TS, into their safety-related motor starting and running
calculations. The inspectors assessed the finding using the SDP and determined
that the finding was of very low safety significance (Green) because the finding
represented a design deficiency confirmed not to result in the loss of functionality of
Enclosure
4
safety-related loads due to the availability of related transformer load tap changers
(LTCs) that were installed to improve a degraded voltage condition.
The inspectors reviewed the performance deficiency for cross-cutting aspects and
determined that none were applicable since this performance deficiency was not
indicative of current licensee performance as the design calculation discussed above
was not recently performed. (Section 4OA5.3)
B. Licensee-Identified Violations
None
Enclosure
REPORT DETAILS
Summary of Plant Status
Unit 1 operated at or near 100 percent rated thermal power (RTP) until November 14, 2010,
when the A Main Bank Transformer alarmed due to a loss of control power to the cooling fans
and pumps resulting in uncontrolled increase in winding temperatures necessitating a manual
Rx Trip. The unit was returned to full power operation on November 19, 2010. The unit
operated at or near 100 percent RTP for the remainder of the inspection period.
1. REACTOR SAFETY
Cornerstones: Initiating Events, Mitigating Systems, and Barrier Integrity
1R01 Adverse Weather Protection
Readiness for Seasonal Extreme Weather Readiness
a. Inspection Scope
The inspectors reviewed licensee actions taken in preparation for low temperature
weather conditions to limit the risk of freeze-related initiating events and to adequately
protect mitigating systems from its effects. The inspectors reviewed licensee procedure
1-PI-OPS-1-FP, Freeze Protection, and walked down selected components associated
with the five areas listed below to evaluate implementation of plant freeze protection,
including the material condition of insulation, heat trace elements, and temporary heated
enclosures. Corrective actions for items identified in relevant problem evaluation reports
(PERs) and work orders (WOs) were assessed for effectiveness and timeliness. This
inspection satisfied one inspection sample for extreme weather readiness. Documents
reviewed are listed in the attachment to this report.
- Refueling water storage tank (RWST) freeze protection preparations
- A-train and B-train essential raw cooling water (ERCW) system freeze protection
preparations
- A-train and B-train high pressure fire protection system freeze protection
preparations
- Main feedwater sensing lines freeze protection preparations
- Diesel generator building freeze protection preparations
b. Findings
No findings were identified.
Enclosure
6
1R04 Equipment Alignment
Partial System Walkdowns
a. Inspection Scope
The inspectors conducted three equipment alignment partial walkdowns, listed below, to
evaluate the operability of selected redundant trains or backup systems with the other
train or system inoperable or out of service. The inspectors reviewed the functional
system descriptions, Updated Final Safety Analysis Report (UFSAR), system operating
procedures, and technical specifications (TS) to determine correct system lineups for the
current plant conditions. The inspectors performed walkdowns of the systems to verify
that critical components were properly aligned and to identify any discrepancies which
could affect operability of the redundant train or backup system. Documents reviewed
are listed in the Attachment.
- Partial walkdown of 1B containment spray (CS) pump following maintenance
activities on 1B CS pump
- Partial walkdown of C-S component cooling system (CCS) pump following
maintenance activities
- 1A motor-driven auxiliary feedwater (MDAFW) pump while B MDAFW pump out of
service (OOS) for maintenance
b. Findings
No findings were identified.
1R05 Fire Protection
Fire Protection Tours
a. Inspection Scope
The inspectors conducted tours of the 10 areas important to reactor safety, listed below,
to verify the licensees implementation of fire protection requirements as described in the
Fire Protection Program, Standard Programs and Processes (SPP)-10.0, Control of Fire
Protection Impairments, NPG-SPP-18.4.7, Control of Transient Combustibles, NPG-
SPP-18.4.8, Control of Ignition Sources (Hot Work). The inspectors evaluated, as
appropriate, conditions related to: (1) licensee control of transient combustibles and
ignition sources; (2) the material condition, operational status, and operational lineup of
fire protection systems, equipment, and features; and (3) the fire barriers used to prevent
fire damage or fire propagation. This activity constituted ten inspection samples.
- Cable Spreading Room
- 480 V RX MOV Board Room 1A
- 480 V RX MOV Board Room 1B
- 480 V RX MOV Board Room 2A
Enclosure
7
- 480 V RX MOV Board Room 2B
- Vital Battery Rooms I, II, III, IV and V
b. Findings
No findings were identified.
.2 Annual Drill Observations
a. Inspection Scope
On November 9, 2010, the inspectors observed an announced fire drill for a simulated
fire of the 6.9 kV Unit Board 1D. The drill was observed to evaluate the readiness of the
plant fire brigade to fight fires. The inspectors verified that the licensee staff identified
deficiencies; openly discussed them in a self-critical manner at the drill debrief, and took
appropriate corrective actions. Specific attributes evaluated were: (1) specified number
of individuals responding; (2) proper wearing of turnout gear; (3) self-contained breathing
apparatus available and properly worn and used; (4) control room personnel followed
procedures for verification and initiation of response; (5) fire brigade leader exhibited
command and had a copy of the pre-fire plan; (6) fire brigade leader maintained control
starting at the dress-out area; (7) fire brigade response timely and followed the
appropriate access route; (8) control/command set up near the location and
communications were established; (9) proper use and layout of fire hoses; (10) fire area
entered in a controlled manner; (11) sufficient firefighting equipment brought to the
scene; (12) search for victims and propagation of the fire into other plant areas; (13)
utilization of pre-planned strategies; (14) adherence to the pre-planned drill scenario and
drill objectives acceptance criteria were met; and (15) firefighting equipment returned to
a condition of readiness to respond to an actual fire. This activity constituted one
inspection sample.
b. Findings
No findings were identified.
1R06 Flood Protection Measures
a. Inspection Scope
The inspectors reviewed internal flood protection measures for the intake pumping
station flood protection features. The features were examined to verify that they were
installed and maintained consistent with the plant design basis. The inspectors also
reviewed the licensees flooding study calculation for determining maximum flood level
in all building rooms for piping failures in both the essential raw cooling water (ERCW)
system and the fire protection system. The inspectors confirmed that flood mitigation
features such as drains and curbs were not degraded in such a manner as to adversely
impact the conclusions of the study. Documents reviewed are listed in the attachment
to this report. This inspection satisfied one inspection sample.
Enclosure
8
b. Findings
No findings were identified.
1R07 Heat Sink Performance
a. Inspection Scope
The inspectors performed two heat sink performance reviews. The inspectors reviewed
the licensees program for maintenance and testing of the 1A-A emergency diesel
generator (EDG) heat exchangers. Specifically, the review included the performance
testing and analysis of the 1A1 (1-HTX-082-720B1) and 1A2 (1-HTX-082-720B2) EDG
jacket water heat exchangers. The inspectors reviewed the ERCW system description,
the heat exchanger performance, and the eddy current testing program document as
well as completed WOs documenting the testing and visual inspection and associated
corrective actions to verify that corrosion or fouling did not impact the heat exchanger
from achieving its design basis heat removal capacity. The inspectors reviewed periodic
test data of ERCW flow rates as well as inlet and outlet temperatures to determine
whether potential degradations were being monitored and/or prevented. The inspectors
also reviewed eddy current inspection results to determine whether wall loss indications
and tube plugging requirements were being identified. The inspectors reviewed the
fouling factor calculation. Documents reviewed are listed in the attachment to this
report. This inspection satisfied two annual inspection samples.
b. Findings
No findings were identified.
1R11 Licensed Operator Requalification
.1 Quarterly Review
a. Inspection Scope
On November 24, 2010, the inspectors observed the annual simulator examination of
Operations Crew 2 conducted per 3-OT-SRE0004A, Feed Water Isolation Followed by a
Steam Generator Tube Rupture, Revision 5. The plant conditions led to an Alert level
classification. Also observed was 3-OT-SRE0032, Loss of Coolant Accident from 75%
Power, Revision 4. The plant conditions led to an Alert level classification. Performance
Indicator credit was taken.
The inspectors specifically evaluated the following attributes related to the operating
crews performance:
- Clarity and formality of communication
- Ability to take timely action to safely control the unit
- Prioritization, interpretation, and verification of alarms
Enclosure
9
- Correct use and implementation of abnormal operating instructions (AOIs), and
emergency operating instructions (EOIs)
- Timely and appropriate Emergency Action Level declarations per Emergency Plan
Implementing Procedures (EPIP)
- Control board operation and manipulation, including high-risk operator actions
- Command and Control provided by the unit supervisor and shift manager
The inspectors attended the post exam critique to assess the effectiveness of the
licensee evaluators and to verify that performance issues identified by the evaluators
were comparable to issues identified by the inspector.
b. Findings
No findings were identified.
.2 Annual Written Test Review
a. Inspection Scope
December 17, 2010, the licensee completed the comprehensive biennial requalification
written examinations and annual requalification operating tests required to be
administered to all licensed operators in accordance with 10 CFR 55.59(a)(2). The
inspectors performed an in-office review of the overall pass/fail results of the written
examinations, individual operating tests and the crew simulator operating tests. These
results were compared to the thresholds established in Manual Chapter 609 Appendix I,
Operator Requalification Human Performance Significance Determination Process.
b. Findings
No findings were identified.
.3 Biennial Inspection
a. Inspection Scope
The inspectors reviewed the facility operating history and associated documents in
preparation for this inspection. During the week of November 15, 2010, the inspectors
reviewed documentation, interviewed licensee personnel, and observed the
administration of operating tests associated with the licensees operator requalification
program. Each of the activities performed by the inspectors was done to assess the
effectiveness of the facility licensee in implementing requalification requirements
identified in 10 CFR Part 55, Operators Licenses. The evaluations were also
performed to determine if the licensee effectively implemented operator requalification
guidelines established in NUREG-1021, Operator Licensing Examination Standards for
Power Reactors, and Inspection Procedure 71111.11, Licensed Operator
Requalification Program. The inspectors also evaluated the licensees simulation
facility for adequacy for use in operator licensing examinations using ANSI/ANS-3.5
Enclosure
10
1988, American National Standard for Nuclear Power Plant Simulators for use in
Operator Training and Examination. The inspectors also reviewed Unit 2 Job
Familiarization Guides associated with system familiarization for Unit 2 construction.
The inspectors observed two crews during the performance of the operating tests.
Documentation reviewed included written examinations, Job Performance Measures
(JPMs), simulator scenarios, licensee procedures, on-shift records, simulator
modification request records, simulator performance test records, operator feedback
records, licensed operator qualification records, remediation plans, watchstanding
records, and medical records. The records were inspected using the criteria listed in
Inspection Procedure 71111.11. Documents reviewed during the inspection are listed in
the Attachment.
b. Findings
No findings were identified.
1R12 Maintenance Effectiveness
a. Inspection Scope
The inspectors reviewed the two performance-based problems listed below. A review
was performed to assess the effectiveness of maintenance efforts that apply to scoped
structures, systems, or components (SSCs) and to verify that the licensee was following
the requirements of TI-119, Maintenance Rule Performance Indicator Monitoring,
Trending, and Reporting 10 CFR 50.65, and SPP-6.6, Maintenance Rule Performance
Indicator Monitoring, Trending, and Reporting 10 CFR 50.65. Reviews focused, as
appropriate, on: (1) appropriate work practices; (2) identification and resolution of
common cause failures; (3) scoping in accordance with 10 CFR 50.65; (4)
characterization of reliability issues; (5) charging unavailability time; (6) trending key
parameters; (7) 10 CFR 50.65 (a)(1) or (a)(2) classification and reclassification; and (8)
the appropriateness of performance criteria for SSCs classified as (a)(2) or goals and
corrective actions for SSCs classified as (a)(1).
- Review of the Eighth Periodic Summary Assessment Report (A3)
- Review of 10 CFR 50.65 (a)(1) plan for the ice condenser following icing issues
b. Findings
Introduction. A Green, non-cited violation of 10 CFR 50.65(a)(2), was identified by the
inspectors for the licensees failure to set goals and monitor the performance and
condition of the B Main Control Room (MCR) Air Conditioning system as required by
10 CFR 50.65(a)(1), and had no justification for not doing so, after it had failed to
demonstrate effective control of the performance or condition of the system through
appropriate preventive maintenance. Per 10 CFR 50.65(a)(2), effective control of SSC
performance and condition through appropriate preventive maintenance must be
demonstrated in order for the monitoring under Paragraph (a)(1) not to be required.
Therefore, a non-cited violation of 10 CFR 50.65(a)(2) was identified.
Enclosure
11
Description. The inspectors reviewed CDEs related to the B MCR Air Conditioning
Train and questioned whether three system failures were actually functional failures as
defined by the licensees procedures. Two of these failures were related to a cooling
water temperature control valve sticking open, causing an interruption of cooling water
flow, rendering the chiller inoperable. The third was related to the chiller tripping during
a fast bus transfer, also rendering the chiller inoperable. The licensee had initially
concluded that these were not functional failures.
Inspectors interviewed the system engineer, engineering supervision, and the
maintenance rule coordinator, questioning the analysis of the three CDEs that had been
classified as non-functional failures. Following the inspectors questions, the licensee
performed a re-evaluation of the CDEs in question, which included benchmarking with
other utilities, and determined the three CDEs should have been classified as functional
failures. The performance criterion established in licensee procedure TI-119, was no
more than three functional failures, per train, within a 24 month interval. The inspectors
determined that the addition of these three functional failures to the one existing
functional failure caused the performance criterion of TI-119 to be exceeded. The
maintenance rule expert panel re-evaluated the performance of the B MCR Air
Conditioning Train for movement from maintenance rule category a(2) to category a(1)
and determined that category a(1) was the appropriate classification.
The inspectors determined that the improper classification of the system functional
failures that ultimately led to the system being move into an a(1) monitoring status
constituted a failure by the licensee to demonstrate that the performance or condition of
the B Main Control Room (MCR) Air Conditioning system had been effectively controlled
through the performance of appropriate scheduled maintenance.
Analysis. The licensees failure to demonstrate that the performance or condition of the
B Main Control Room (MCR) Air Conditioning system had been effectively controlled
through the performance of appropriate scheduled maintenance (10 CFR 50.65(a)(2))
without implementing goal setting and monitoring requirements of 50.65(a)(1), was
determined to be a performance deficiency. The inspectors determined that this
performance deficiency was more than minor since the B MCR Air Conditioning Train
was not placed in 50.65(a)(1) monitoring status in a timely manner which if left
uncorrected, could become a more significant safety concern.
The inspectors determined this finding to have very low safety significance (Green)
because it was not among the contributing causes of the degraded performance and the
condition of the B Main Control Room (MCR) Air Conditioning system. The cause of the
finding was directly related to the cross-cutting area of Problem Identification and
Resolution, evaluation aspect of the corrective action program component, in that, the
licensee failed to thoroughly evaluate failures and determine those failures to be
functional failures of the B MCR Air Conditioning System such that the system was
placed in category a(1) in a timely manner. P.1(c)
Enforcement. 10 CFR 50.65(a)(1) requires, in part, that licensees shall monitor the
performance or condition of system, structures and components within the scope of the
rule against licensee-established goals in a manner sufficient to provide reasonable
Enclosure
12
assurance the system, structures and components are capable of fulfilling their intended
safety functions. 10 CFR 50.65(a)(2) requires, in part, that the monitoring specified in
paragraph (a)(1) is not required where it has been demonstrated the performance or
condition of a system, structures and components is being effectively controlled through
the performance of appropriate preventive maintenance such that the system, structures
and components remains capable of performing its intended function.
Contrary to the above, the licensee failed to satisfy the requirements of 10 CFR
50.65(a)(2), to demonstrate that the performance or condition of the B MCR Air
Conditioning Train system had been effectively controlled through the performance of
appropriate scheduled maintenance and subsequently failed to implement monitoring of
the system against licensee-established goals as required by 10 CFR 50.65(a)(1).
Specifically, the licensee failed to identify and properly account for three functional
failures which demonstrated that the performance of the system was not being
effectively controlled and, as a result, goal setting and monitoring, as required by 10
CFR 50.65(a)(1), was required since October 9, 2009, but not initiated or performed.
The licensee implemented goal setting and monitoring as described in 50.65 (a)(1) for
the B MCR Air Conditioning Train on October 21, 2010. Because this inspection finding
was characterized as having very low risk significance (Green) and has been entered in
the licensees corrective action program as PER205438, this violation is being treated as
a non-cited violation, consistent with the NRC Enforcement Policy: NCV 05000390/2010005-01, Failure to Monitor Performance of the B MCR Air Conditioning
Train.
1R13 Maintenance Risk Assessments and Emergent Work Control
a. Inspection Scope
The inspectors evaluated, as appropriate, for the four work activities listed below: (1) the
effectiveness of the risk assessments performed before maintenance activities were
conducted; (2) the management of risk; (3) that, upon identification of an unforeseen
situation, necessary steps were taken to plan and control the resulting emergent work
activities; and (4) that maintenance risk assessments and emergent work problems were
adequately identified and resolved. The inspectors verified that the licensee was
complying with the requirements of 10 CFR 50.65 (a)(4); SPP-7.0, Work Control and
Outage Management; NPG-SPP-07.1, One Line Work Management; and TI-124,
Equipment to Plant Risk Matrix. This inspection satisfied four inspection samples for
Maintenance Risk Assessment and Emergent Work Control.
- Risk assessment for emergent failure of 1B main control room (MCR) chiller during
A-train work week
- Risk assessment for work week 605
- Risk assessment of 1A motor-driven auxiliary feedwater (MDAFW) pump component
Enclosure
13
b. Findings
No findings were identified.
1R15 Operability Evaluations
a. Inspection Scope
The inspectors reviewed two operability evaluations affecting risk-significant mitigating
systems, listed below, to assess, as appropriate: (1) the technical adequacy of the
evaluations; (2) whether continued system operability was warranted; (3) whether the
compensatory measures, if involved, were in place, would work as intended, and were
appropriately controlled; (4) where continued operability was considered unjustified, the
impact on TS Limiting Conditions for Operation (LCOs) and the risk significance in
accordance with the significant determination process (SDP). The inspectors verified
that the operability evaluations were performed in accordance with NPG-SPP-03.1,
Corrective Action Program. Documents reviewed are listed in the Attachment.
- Daily ice removal from ice condenser intermediate deck doors
- FCV-061-193A ice condenser isolation valve AO contact stuck
b. Findings
No findings were identified.
1R19 Post-Maintenance Testing
a. Inspection Scope
The inspectors reviewed five post-maintenance test procedures and/or test activities,
(listed below) as appropriate, for selected risk-significant mitigating systems to assess
whether: (1) the effect of testing on the plant had been adequately addressed by control
room and/or engineering personnel; (2) testing was adequate for the maintenance
performed; (3) acceptance criteria were clear and adequately demonstrated operational
readiness consistent with design and licensing basis documents; (4) test instrumentation
had current calibrations, range, and accuracy consistent with the application; (5) tests
were performed as written with applicable prerequisites satisfied; (6) jumpers installed or
leads lifted were properly controlled; (7) test equipment was removed following testing;
and (8) equipment was returned to the status required to perform its safety function. The
inspectors verified that these activities were performed in accordance with SPP-8.0,
Testing Programs; NPG-SPP-06.3, Pre-/Post-Maintenance Testing; and NPG-SPP-07.1,
On Line Work Management.
- WO 10-813997-000, 1-FCV-77-19, RCDT to vent HDR flow control valve stroke time
- WO 09-821944, 1-MVOP-077-0010, RCDT pump discharge valve replacement
Enclosure
14
- WO 111516647, 1B MDAFW 1-FCV-3-132 maintenance
- WO 111238674, Replacement of ERCW pump D-A
b. Findings
No findings were identified
1R22 Surveillance Testing
a. Inspection Scope
The inspectors witnessed seven surveillance tests and/or reviewed test data of selected
risk-significant SSCs, listed below, to assess, as appropriate, whether the SSCs met the
requirements of the TS; the UFSAR; SPP-8.0, Testing Programs; NPG-SPP-06.9.2,
Surveillance Test Program; and SPP-9.1, ASME Section XI. The inspectors also
determined whether the testing effectively demonstrated that the SSCs were
operationally ready and capable of performing their intended safety functions.
In-Service Test:
Performance Test
- WO 10-814970-000, 1-SI-72-901-B, Containment spray pump 1B-B quarterly
performance test
- WO 10-814988-000, 1-SI-31-901-B, Quarterly valve full stroke exercising during
plant operation chilled water - B-train
Containment Isolation Valve Leak Rate:
- WO 10-814987-000, 1-SI-30-701, Containment isolation valve local leakrate test -
purge air
Other Surveillances
- WO 10-815229-000, Monthly Diesel Generator Start and Load Test (1B)
- WO 111539446, 1-SI-0-24, Measurement of At Power Moderator Temperature
Coefficient
b. Findings
No findings were identified.
Enclosure
15
Cornerstone: Emergency Preparedness
1EP6 Drill Evaluation
a. Inspection Scope
On October 7, 2010, the inspectors observed a licensee-evaluated emergency
preparedness drill, listed below, to verify that the emergency response organization was
properly classifying the event in accordance with EPIP-1, Emergency Plan Classification
Flowchart, and making accurate and timely notifications and protective action
recommendations in accordance with EPIP-2, Notification of Unusual Event; EPIP-3,
Alert; EIPIP-4, Site Area Emergency; EPIP-5, General Emergency; and the Radiological
Emergency Plan. In addition, the inspectors verified that licensee evaluators were
identifying deficiencies and properly dispositioning performance against the performance
indicator criteria in Nuclear Energy Institute (NEI) 99-02, Regulatory Assessment
Performance Indicator Guideline.
- A steam generator tube rupture leads to an Alert classification
- A PORV on the ruptured steam generator fails open, requiring Site Area Emergency
classification
b. Findings
No findings were identified.
4. OTHER ACTIVITIES
4OA2 Identification & Resolution of Problems
.1 Review of Items Entered into the Corrective Action Program (CAP)
As required by Inspection Procedure 71152, Identification and Resolution of Problems,
and in order to help identify repetitive equipment failures or specific human performance
issues for follow-up, the inspectors performed a daily screening of items entered into the
licensees CAP. This review was accomplished by reviewing daily PER summary
reports and attending daily PER review meetings.
.2 Semi-Annual Review to Identify Trends
a. Inspection Scope
As required by IP 71152, Identification and Resolution of Problems, the inspectors
performed a review of the licensees CAP and associated documents to identify trends
that could indicate the existence of a more significant safety issue. The inspectors
review was focused on human performance trends, licensee trending efforts, and
repetitive equipment and corrective maintenance issues. The inspectors also
considered the results of the daily inspector CAP item screening discussed in Section
4OA2.1. The inspectors review nominally considered the six-month period of July 2010
Enclosure
16
through December 2010, although some examples expanded beyond those dates when
the scope of the trend warranted.
b. Observations
No findings were identified. However, the inspectors identified a number of instances
where the PER screening committees (PSC) review of incoming PERs failed to
recognize conditions adverse to quality which required potential operability reviews,
potential reportablity reviews, or the need to upgrade some PER classifications. Also,
examples of degraded or non-conforming conditions of plant equipment related to the
current licensing basis were not addressed by the PSC. Inspectors noted a trend in the
number of instances where questioning from the inspectors was necessary for the
licensee to address these types of issues. The inspectors discussed these issues with
the licensee during the exit meeting and the licensee entered them into the corrective
action program as PERs 252780, 252215 and 241755.
.3 Annual Sample: Corrective actions associated with NCV 05000390/2008005-01, Failure
to Translate ERCW Pump Coupling Material Change into Procedures
a. Inspection Scope
The inspectors reviewed the plan and implementation of corrective actions for non-cited
violation (NCV)05000390/2008005-01, which were documented in PER 148716.
b. Findings and Observations
The corrective action plan for PER 148716 implemented DCN 52920 to replace all
ERCW pumps w/ pumps capable of 2 unit operation. This combined with changes to MI-
67.1, Removal, Inspection, And Repair Of Essential Raw Cooling Water Pumps,
changed all existing 410 Stainless Steel ERCW pump shaft couplings with XM-19 alloy
shaft couplings. The inspectors reviewed replacement work orders and the licensees
extent of cause and condition. The licensee determined during an the extent of
condition review that the Screen Wash and High Pressure Fire Pumps could have the
same susceptibility and pursuing potential design changes for these components. The
licensee also determined that a weakness existed in follow-up of NRC Information
Notices.
No findings were identified.
4OA3 Event Follow-up
a. Inspection Scope
On November 14, 2010, Unit 1 reactor was manually tripped as a result of the A Main
Bank Transformer alarming due to a loss of control power to the cooling fans and pumps
resulting in a loss of oil cooling which resulted in an uncontrolled increase in the
transformers winding temperatures. All systems/components behaved as expected
Enclosure
17
except the #1 main feedwater bypass valve isolation which indicated mid-position. This
was later determined to be a limit switch issue and the valve was actually shut.
Inspectors responded to the event, reviewed plant logs, procedures, and corrective
action documents. The inspectors interviewed personnel associated with the reactor trip
and abnormal transformer indications.
b. Findings
No findings were identified.
4OA5 Other Activities
.1 Quarterly Resident Inspector Observations of Security Personnel and Activities
a. Inspection Scope
During the inspection period, the inspectors conducted observations of security force
personnel and activities to ensure that the activities were consistent with licensee
security procedures and regulatory requirements relating to nuclear plant security.
These observations took place during both normal and off-normal plant working hours.
These quarterly resident inspector observations of security force personnel and activities
did not constitute any additional inspection samples. Rather, they were considered an
integral part of the inspectors normal plant status review and inspection activities.
b. Findings
No findings were identified.
(Closed) Reactor Coolant System Dissimilar Metal Butt Welds (TI 2515/172, Revision 1)
a. Inspection Scope
The inspectors conducted a review of the licensees activities regarding licensee
dissimilar metal butt weld (DMBW) mitigation and inspection implemented in accordance
with the industry self-imposed mandatory requirements of Materials Reliability Program
(MRP) 139, Primary System Piping Butt Weld Inspection and Evaluation Guidelines.
Temporary Instruction (TI) 2515/172, Reactor Coolant System Dissimilar Metal Butt
Welds, Revision 1, was issued May 27, 2010, to support the evaluation of the licensees
implementation of MRP-139.
On December 8, 2010, the inspectors performed a review in accordance with TI
2515/172, Revision 1, as described in the Observations section below:
Enclosure
18
b. Observations
The licensee has met the MRP-139 deadlines for baseline examinations of all welds
scoped into the MRP-139 program. TI 2515/172, Revision 1, is considered closed. In
accordance with requirements of TI 2515/172, Revision 1, the inspectors evaluated the
following areas:
(1) Implementation of the MRP-139 Baseline Inspections
This portion of the TI was not inspected during the period of this inspection report but
was previously covered in NRC Inspection Report 05000390/2008003.
(2) Volumetric Examinations
This portion of the TI was not inspected during the period of this inspection report, but
was previously covered in NRC Inspection Report 05000390/2010002.
(3) Weld Overlays
There were no weld overlay activities performed or planned by this licensee to comply
with their MRP-139 commitments.
(4) Mechanical Stress Improvement (SI)
This portion of the TI was not inspected during the period of this inspection report, but
was previously covered in NRC Inspection Report 05000390/2008003.
(5) Application of Weld Cladding and Inlays
There were no weld cladding nor inlay activities performed or planned by this licensee to
comply with their MRP-139 commitments.
(6) Inservice Inspection Program
This portion of the TI was not inspected during the period of this inspection report, but
was previously covered in NRC Inspection Report 05000390/2008003.
c. Findings
No findings were identified.
.2 (Closed) URI 05000390/2009002-003: Acceptability of Seismic Qualification of 120VAC
Vital Instrumentation Board Circuit Breakers
a. Inspection Scope
During the 2009 Evaluations of Changes, Tests, or Experiments and Permanent Plant
Modifications inspection, an unresolved item was indentified related to the adequacy of
seismic qualification of station 120VAC vital instrumentation boards. The inspectors
Enclosure
19
were concerned that the breaker mounting did not adequately represent the plant-
specific mounting and that the breakers were not tested at adequate accelerations to
fully bound the required response spectrum (RRS) across the ground frequency range.
The item was unresolved pending further review of the adequacy of the licensees
seismic qualification of the installed equipment.
b. Findings
Introduction: A green NCV of 10 CFR 50, Appendix B, Criterion III, Design Control, was
identified for the failure to assure that appropriate quality standards were specified and
included in design documents and that deviations from such standards were controlled.
Specifically, the licensee failed to ensure that the substitute Heinemann Circuit Breakers
utilized in the station 120VAC vital instrumentation boards were properly seismically
qualified for their application.
Description: The licensee originally procured the 120VAC vital instrumentation boards
as a complete functional unit, dedicated and seismically qualified by the vendor. In the
early 1990s, the licensee implemented a complete replacement of the Heinemann
circuit breakers in the instrumentation boards with commercial grade breakers from the
same manufacturer. A different third party vendor was contracted to perform seismic
qualifications for the replacement breakers.
Both, the licensee and the contract vendor, committed to IEEE Standard 344 (1975),
which requires, in part, that the test mounting dynamically simulate the plant-specific
mounting and that the test accelerations adequately bound the required response
spectrum (RRS) for the application.
Given limited accelerometer mounting locations on the original 1974 qualification testing,
the licensee translated maximum accelerations seen on the panel itself as bounding the
subcomponent accelerations without adequately demonstrating the rigidity of mounting
necessary to support that assumption. As the mounting configuration of the devices to
the test platform did not mimic the actual installed mounting, the licensee had
responsibility to ensure, by analysis, that the test accelerations adequately bounded the
RRS. The licensee failed to ensure such analysis was conducted. Specifically,
calculation WCG-ACQ-1004 failed to fully establish that the method of support of the
breakers within the board was a rigid mounting system, that the 1992 test mounting
represents a suitable mounting method, or that the test accelerations to which the device
was subjected were, in fact, bounding.
In October 2010, the licensee issued calculation WCG-ACQ-1301, Frequency Evaluation
of the Heinemann Breaker Support Structure, Rev. 000 to demonstrate the rigidity of the
breaker mounting system by performing a finite element analysis of the panel front plate
and rear angle supports used for impinging the breakers to satisfy the expectation of
rigidity. Calculation WCG-ACQ-1004 was revised (Revision 2) to credit calculation
WCG-ACQ-1301 with that demonstration to justify the ability to perform seismic testing
on an individual component basis, to investigate the potential for local structural support
Enclosure
20
flexibility and associated amplifications, and to demonstrate the appropriateness of the
3G test level used in the 1992 qualification testing.
Additionally, at the time of inspection in March 2009, the licensee initiated PER 165130
to enhance existing work instructions to specify the tightness requirement of press-fit
devices on various boards.
The licensee presented all of these details in a public meeting held on December 16,
2010, intended to address NOV 05000391/2010603-08 associated with the Unit 2
Completion Projects acceptance and application of the new breakers (identified in URI
05000390/2009002-03) based on the 1992 testing in question. The inspectors
determined that the licensee response was inadequate in that it did not demonstrate that
the 1992 test adequately represented the installed configuration and in that the snug fit
configuration cannot be adequately assured through the maintenance and testing
procedures as presented.
Analysis: Failure to adequately qualify commercial-grade molded-case circuit breakers
to their safety-related application is a performance deficiency. This performance
deficiency is more than minor because it affected the design control attribute of the
mitigating systems cornerstone objective to ensure the availability, reliability, and
capability of systems that respond to initiating events to prevent undesirable
consequences. Specifically, adequate measures were not implemented to ensure the
station 120VAC vital instrumentation boards had proper seismic qualification for their
application. The inspector assessed this finding for significance in accordance with NRC
Manual Chapter 0609, Appendix A, Attachment 1, Significance Determination Process
(SDP) for Reactor Inspection Findings for At-Power Situations, and determined that it
was of very low safety significance (Green) as the devices in question had been
intrinsically qualified for this application as part of a complete panel test by the original
vendor. This finding was reviewed for cross-cutting aspects and none were identified as
it was determined to not reflect current licensee performance.
Enforcement: 10 CFR 50, Appendix B, Criterion III, Design Control states, in part, that
design control measures shall assure that appropriate quality standards are specified
and included in design documents and that deviations from such standards are
controlled. Contrary to the above, the licensee failed to demonstrate the necessary
conditions for the commercial grade dedication and seismic qualification of molded case
circuit breakers to safety-related application within the station 120VAC vital
instrumentation boards. This condition existed since commercial operations began in
1995. This finding was entered into the licensees corrective action program as PER
171695 related to the URI. Because the finding was of very low safety significance and
has been entered into the licensee's corrective action program, this violation is being
treated as a non-cited violation (NCV), consistent with the NRC Enforcement Policy:
NCV 05000390/2010005-XX, Failure to Adequately Qualify Molded-Case Circuit
Breakers to Safety-Related Application Through Commercial Grade Dedication.
Enclosure
21
.3 (Closed) URI 05000390/2010008-02, Worst Case 6900 VAC Bus Voltage in Design
Calculations
Introduction: The NRC identified a Green non-cited violation (NCV) of 10 CFR 50,
Appendix B, Criterion III, Design Control, for the failure to correctly translate the 6900
VAC emergency bus undervoltage trip value specified in Technical Specifications (TS)
into design calculations for motor starting and loading. The values used by the licensee
in the design calculations were non-conservative with respect to the specified TS values.
This issue was initially discussed as URI 05000390/2010008-02: Worst Case 6900
VAC Bus Voltage in Design Calculations.
Description: Offsite power at Watts Bar is normally provided to the Class 1E 6900 VAC
buses from the 161 kV offsite power system through the Common Station Service
Transformers (CSSTs). Watts Bar TS Section 3.3.5-1, item 2, Loss of Power Diesel
Generator Start Instrumentation, requires and specifies the undervoltage and degraded
voltage relay trip setpoints, including allowable values and time delays associated with
the safety-related 6900 VAC buses. These degraded voltage setpoints provide the
bases for the minimum voltage available to all safety-related equipment such as motors,
contactors, and solenoid valves during a postulated degraded voltage scenario.
At Watts Bar, the degraded voltage relays initiate the nominal 10 second time delay at
the TS specified relay voltage setting. When the 10 second time delay has elapsed, the
plant loads are removed from the offsite power supply and transferred to the onsite
emergency diesel generators. The degraded voltage relays drop-out (de-energize)
when sufficient voltage is not available and normally pick-up (energize) if voltage is
recovered within the 10 second delay on the 6900 VAC bus. The degraded voltage
relay settings at Watts Bar are in accordance with TS Table 3.3.5-1 which states the
values to be as follows: Allowable Value 6570 VAC, Trip Setpoint between 6606 VAC
and 6593 VAC.
The inspector reviewed licensee calculation of record WBN-EEB-MS-TI-06-0029,
Degraded Voltage Analysis, Rev. 31, which evaluated motor starting voltages at the
beginning of a design basis loss of coolant accident (LOCA) concurrent with a degraded
grid condition. This calculation used the degraded voltage setpoint of 6672 V to analyze
post LOCA load motor starting. This voltage of 6672 VAC used in the calculation was
non-conservative with respect to the voltage specified in TS which specified a maximum
value of 6606 VAC.
Analysis: The failure to use the degraded voltage relay setpoint values as specified in
TS and installed in the plant for the 6900 VAC bus electrical design calculation was a
performance deficiency. This finding is more than minor because it affects the Design
Control attribute of the Mitigating Systems Cornerstone. It impacts the cornerstone
objective of ensuring the availability, reliability, and operability of the 6900 VAC safety
buses to perform the intended safety function during a design basis event. The potential
availability, reliability, and operability of the 6900 VAC safety buses during a potential
degraded voltage condition was impacted as the licensee calculation used a non
conservative degraded voltage input, with respect to the values specified in TS, into their
safety-related motor starting and running calculations. The inspectors assessed the
Enclosure
22
finding using the SDP and determined that the finding was of very low safety significance
(Green) because the finding represented a design deficiency confirmed not to result in
the loss of functionality of safety-related loads due to the availability of load tap changers
(LTCs) that are installed to improve a degraded voltage condition.
The inspectors reviewed the performance deficiency for cross-cutting aspects and
determined that none were applicable since this performance deficiency was not
indicative of current licensee performance as the design calculation discussed above
was not recently performed.
Enforcement: 10 CFR 50, Appendix B, Criterion III, Design Control, states, in part, that
measures shall be established to assure that applicable regulatory requirements and the
design basis for structures, systems, and components are correctly translated into
specifications, drawings, procedures, and instructions. This appendix also states in part
that measures shall be established for the selection and review for suitability of
application of processes that are essential to the safety-related functions of the
structures, systems, and components. Watts Bar TS Section 3.3.5-1, Loss of Power
Diesel Generator Start instrumentation, table 3.3.5-1, item 2 specifies the 6900 VC
emergency bus undervoltage (degraded) relay trip setpoints to be as follows: Allowable
Value, 6570 VAC, Trip Setpoint, 6606 VAC and 6593 VAC.
Contrary to the above, since at least December 2001, the licensee failed to assure that
applicable regulatory requirements for undervoltage (degraded) voltage protection,
including those prescribed in TS 3.3.5-1, item 2, were correctly translated into design
calculation, WBN-EEB-MS-TI-06-0029, Degraded Voltage Analysis, Revision 31, which
evaluated motor starting voltages at the beginning of a design basis loss of coolant
accident (LOCA) concurrent with a degraded grid condition. Further, the process used
by the licensee for the selection of input voltage value in the design calculation was non-
conservative with respect to the TS. Specifically, the licensee used the input value of
6672 VAC which was higher than the maximum value of 6606 VAC specified in TS. This
did not result in a loss of function of safety-related loads.
Because this finding is of very low safety significance and was entered into the
licensees corrective action program as PER 296306 this violation is being treated as a
NCV, consistent with the NRC Enforcement Policy. This finding is identified as NCV
05000390, 2010005-:Failure to Use Worst Case 6900 VAC Bus Voltage in Design
Calculations. URI 05000390/2010008-02,Worst Case 6900 VAC Bus Voltage in
Design Calculations is closed.
Enclosure
23
4OA6 Meetings, including Exit
.1 Exit Meeting Summary
An exit meeting was conducted on November 19, 2010, to discuss the findings of the
biennial requalification inspection. The inspectors confirmed that no proprietary
information was reviewed during this inspection.
An interim exit was conducted on December 16, 2010, to discuss the findings associated
with the URI follow-up inspection. Although proprietary information was reviewed during
the inspection, no proprietary information is included in this report.
On January 10, 2011, the inspectors presented the inspection results to Mr. Don
Grissette, Site Vice President, and other members of the licensee staff. The inspectors
confirmed that none of the potential report input discussed was considered proprietary.
4OA7 Licensee Indentified Violations
None
Enclosure
SUPPLEMENTAL INFORMATION
KEY POINTS OF CONTACT
Licensee personnel
G. Boerschig, Plant Manager
M. Brandon, Director, Safety & Licensing (Interim)
J. Bushnell, Licensing Engineer
R. Crews, Operations Training Manager
J. Dalton, Initial Licensing Operator Training Supervisor
T. Detchemende, Emergency Preparedness Manager
B. Ennis, Electrical Engineering
N. Good, Simulator Manager
D. Grissette, Site Vice President
W. Hooks, Radiation Protection Manager
D. Hughes, Training Supervisor
B. Hunt, Operations Superintendent
D. Hutchinson, Chemistry Manager
G. Mauldin, Director, Engineering
M. McFadden, Operations Manager
J. Milner, Technical Support Superintendent, Radiation Protection
D. Murphy, Maintenance Manager (Interim)
M. Pope, Licensing Engineer
C. Riedl, Licensing Manager (Interim)
A. Scales, Work Control Manager
M. Schmader, Training Supervisor
J. Smith, Health Physics Supervisor
W. Thompson, Site Training Director
D. Voeller, Director, Project Management
J. Wilcox, Security Manager
ITEMS OPENED, CLOSED, AND DISCUSSED
Opened
None
Opened and Closed
05000390/2010005-01 NCV Failure to Adequately Monitor the Performance of
the B MCR Air Conditioning Train Under 10 CFR
50.65.05000390/2010005-02 NCV Failure to Adequately Qualify Molded-Case Circuit
Breakers to Safety-Related Application Through
Commercial Grade Dedication. (Section 4OA5.2)
Attachment
2
05000390/2010005-03 NCV Failure to Use Worst Case 6900 VAC Bus Voltage
in Design Calculations. (Section 4OA5.3)
Closed
05000390/2515/172 TI Reactor Coolant System Dissimilar Metal Butt
Welds (Section 4OA5.1)05000390/2009002-03 URI Acceptability of Seismic Qualification of 120VAC
Vital Instrumentation Board Circuit Breakers05000390/2010008-02 URI Worst Case 6900 VAC Bus Voltage in Design
Calculations
Discussed
None
Attachment
LIST OF DOCUMENTS REVIEWED
Section 1R01: Adverse Weather Protection
1-PI-OPS-1-FP, Freeze Protection
PER 272583
Section 1R04: Equipment Alignment
SOI-3.02 Checklist 1, Auxiliary Feedwater System Handswitch Alignment Verification
SOI-3.02 Checklist 2, Auxiliary Feedwater System Electrical Power Alignment Verification
SOI-3.02 Checklist 3, Auxiliary Feedwater System Valve Alignment Verification
SOI-70.01-Attachment 1P, Unit 1 and Common Power Checklist
SOI-7001-Attachment 1V, Unit 1 CCS Normal Power Checklist
SOI-72.01-Attachment 1P, Containment Spray Power Checklist
SOI-72.01-Attachment 1V, Containment Spray Valve Checklist
Section 1R06: Flood Protection Measures
WB-DC-20-28, Intake Pumping Station Watertight Doors at Elevation 722.0
Technical Instruction (TI)-50.023, Intake Pumping Station Strainer Room B Sump Pump A
Performance Test
Technical Instruction (TI)-50.024, Intake Pumping Station Strainer Room B Sump Pump B
Performance Test
TVA Calculation WBN OSG4099 Appendix E, MELB Moderate Energy Line Break (MELB)
Flooding Study (Intake Pumping Station)
WO 10-811526 B Strainer Room Sump Pump B
WO 09-820527 B Strainer Room Sump Pump A
Dwg 1-47610-40
Section 1R07: Heat Sink Performance
TI-79.823 Diesel Generator 2A-A Jacket Water Cooler Performance Test
TI-79.821 Diesel Generator 1A-A Jacket Water Cooler Performance Test
TI-79.000 Program for implementing NRC Generic letter 89.13
Calculation MDQ00008220030077 - DG JWHX
Section 1R11: Licensed Operator Requalification
Job performance measures (JPMs):
JPM 3-OT-JPMR108, Return PRM N-42 to Service Per AOI-4, rev. 3.
JPM 3-OT-JPMR093, Establish RCS Bleed Paths Per FR-H.1, rev. 8.
JPM 3-OT-JPMR018, Perform Boration of the RCS During an ATWS Per FR-S.1., rev. 6.
JPM 3-OT-JPMA049B, 1B-B Diesel Generator Idle Start for Warm Up Per SOI-82.02., rev. 1.
JPM 3-OT-JPMS090A, Classify the Event per the REP (ATWS-Reactor Tripped Locally), rev. 5.
JPM 3-OT-JPMA136, Control the 1B-B Motor-Driven AFW Pump Discharge Pressure Control
Valve Locally per AOI-30.2, Appendix C., rev. 3.
JPM 3-OT-JPMR071A, Align an RHR Train for Hot Leg Recirculation per ES-1.4, rev. 5,
9/1/2010.
JPM 3-OT-JPMR173A, Start Up Upper Containment Purge Per SOI-30.02, rev. 0, 11/01/2010.
JPM 3-OT-JPMR027A, Raise Cold Leg Accumulator Level Per SOI-63.01, rev. 5, 10/05/2010.
JPM 3-OT-JPMS082A, Classify the Event per the REP (Loss of Main Control Room
Annunciation), rev. 8, 10/05/2010.
Attachment
4
Procedures:
OPDP-10, License Status Maintenance, Reactivation and Proficiency for Non-Licensed
Positions, rev. 2, 06/01/2010.
TI-12.10, Control of Sensitive Equipment, rev. 00003, Watts Bar Unit 1.
TRN 11.4, Continuing Training for Licensed Personnel, rev. 0016, 03/11/2010.
TRN 11.8, Operator License Examinations and Renewals, rev. 8, 10/05/2010.
TRN 11.9, Simulator Exercise Guide Development and Revision, rev. 0006, 10/23/2009.
TRN-11.10, Annual Requalification Examination Development and Implementation, rev. 16,
05/26/2010.
TRN-11.12, Job Performance Measure Development, Administration, and Evaluation Manual,
rev. 0004, 07/25/2008.
TRN-11.14, TVA Operator Licensing Examination Security Program, rev. 0004, 07/03/2006.
TRN-12, Simulator Regulatory Requirements, rev. 0009, 10/22/2010.
3TRN-205.2, Evaluation.
Simulator Exam Scenarios (SES):
3-OT-SRE022A, Feedwater Malfunction Followed by Large Break LOCA, rev. 4, 09/29/2010.
3-OT-SRE004A, Feed Water Isolation Followed by a Steam Generator Tube Rupture, rev. 5,
09/30/2010.
Simulator Transient Tests:
Transient Test-2 (TT-2), Loss of Normal and Emergency Feedwater, (2009 and 2010).
TT-4, Simultaneous Four Loop Reactor Coolant Pump Trip, (2009 and 2010).
TT-6, Manual Turbine Trip Without Reactor Trip, (2009 and 2010).
Simulator Steady State Tests:
TRN-12 100%, 75%, 25% Steady-State Performance Test, (2008, 2009, 2010).
Steady State Drift Test60 minute run at 100% power (2010).
Simulator Malfunction Tests:
FW05, Main Feed Pump Trip (2005 and 2009).
FW09, Loss of Vacuum (2003 and 2007).
IA02, Loss of Non-Essential Control Air (2004 and 2008).
IA03, Loss of Essential Control Air (2003 and 2007).
TH09, Fuel Cladding Failure (2003 and 2007).
Written Examinations Reviewed:
Week 2 RO and SRO Biennial Written Exams (2009).
Week 4 RO and SRO Biennial Written Exams (2009).
Week 5 RO and SRO Biennial Written Exams (2009).
Condition Reports:
PER 152195, Unit 1 experienced a reactor trip in response to a turbine trip.
PER 152955, Reactor Trip due to a personnel error - Human Performance.
PER 154635, Human performance - self checking was a flawed defense.
PER 210805, Identifies that SROs are not being trained as ROs to take the OATC position
when it is necessary.
Attachment
5
Other Documents:
Feedback Comments from Licensed Operator Requalification, 2008 to 2010.
Licensed operator medical records (10).
Closed Simulator Discrepancy Reports (DRs) since 2008.
Open/Active Simulator DR List as of 11/15/2010.
Assessment Number - WBN-TRN-10-034, Snapshot Self Assessment Report: Procedure
Adherence and Command and Control issues
2008/2009 Review of LOR Training Program.
3-OT-MSC-147, Self Study Guide, Unit 2 Job Familiarization Guide. (5 Guides)
LER 390/2008-005, Report of Inoperability of Radiation Monitor due to Non-conservative
setpoint.
LER 390/2008-004, Automatic Reactor Trip in Response to Opening of Exciter Field Breaker.
SR 164113, Learning Opportunity (LO) from Licensed Operator Requalification (LOR) Program
Review 2008 and 2009.
SR 164119, Learning Opportunity (LO) from Licensed Operator Requalification (LOR) Program
Review 2008 and 2009. Provide additional training on Logic and Schematic print reading for
the four identified 2009 Biennial Written Exam weakness areas, Steam Dump System,
Containment Isolation Signals, Radiation Monitors, Rod Control System.
Section 1R15: Operability Evaluations
PER 178806
PER 240363
Ice Condenser Trending and Inspection Data, 8/28/2010-10/12/2010
Section 4OA2: Problem Identification and Resolution
PER 148716
MWO 09-816926, ERCW Pump B-A
MWO 05-817978, ERCW Pump A-A
MWO 07-819029, ERCW Pump D-A
MWO 08-822029, ERCW Pump C-A
MWO 09-816921, ERCW Pump E-A
MWO 09-816925, ERCW Pump G-A
MWO 09-816922, ERCW Pump H-A
EDC-53982, Update of ERCW System Description for replaced pumps
DCN 52920, ERCW Pump Replacement
DCN S-1081-A, Shaft and Bearing Material Change
PER 252780 PSC clock reset for missed immediate action to stop missile shield re installation.
PER 252215 PSC clock reset issue was not flagged by PSC as Potential Operability and
Potential Reportability.
PER 241755 - Completeness of actions on pre-startup up PER for Unit 1 related to loose
control board lugs
Attachment
LIST OF ACRONYMS
ANS Alert and Notification System Testing
ARERR Annual Radiological Effluent Release Report
CAP Corrective Action Program
CFR Code of Federal Regulations
CY calendar year
DEP Emergency Response Organization Drill/Exercise Performance
EAL Emergency Action Level
ED electronic dosimeter
ERO Emergency Response Organization
HPT Health Physics Technician
IP Inspection Procedure
LHRA locked high radiation area
LSC liquid scintillation counter
NEI Nuclear Energy Institute
No. Number
NSTS National Source Tracking System
ODCM Offsite Dose Calculation Manual
PCM personnel contamination monitor
PERs Problem Evaluation Report
PI Performance Indicator
PS Planning Standard
QA Quality Assurance
RCA radiologically controlled area
RG Regulatory Guide
REMP Radiological Environmental Monitoring Program
Rev. Revision
RS Radiation Safety
RWP radiation work permit
SAM small article monitor
TBSS Turbine Building System Sump
TI Temporary Instruction
TLDs thermoluminescent dosimeters
TS Technical Specification
UFSAR Updated Final Safety Analysis Report
U1 Unit 1
U2 Unit 2
VHRA very high radiation area
WBC whole body count
Attachment