ML090440261
| ML090440261 | |
| Person / Time | |
|---|---|
| Site: | Watts Bar  |
| Issue date: | 02/12/2009 |
| From: | Heather Gepford Reactor Projects Region 2 Branch 6 |
| To: | Swafford P Tennessee Valley Authority |
| References | |
| IR-08-005 | |
| Download: ML090440261 (38) | |
See also: IR 05000390/2008005
Text
UNITED STATES
NUCLEAR REGULATORY COMMISSION
REGION II
SAM NUNN ATLANTA FEDERAL CENTER
61 FORSYTH STREET, SW, SUITE 23T85
ATLANTA, GEORGIA 30303-8931
February 12, 2009
Mr. Preston D. Swafford
Chief Nuclear Officer and Executive Vice President
Tennessee Valley Authority
3R Lookout Place
1101 Market Street
Chattanooga, TN 37402-2801
SUBJECT:
WATTS BAR NUCLEAR PLANT - NRC INTEGRATED INSPECTION REPORT
05000390/2008005, 05000391/2008005, AND 05000390/2008501 AND
EXERCISE OF ENFORCEMENT DISCRETION
Dear Mr. Swafford:
On December 31, 2008, the United States Nuclear Regulatory Commission (NRC) completed
an inspection at your Watts Bar Nuclear Plant, Units 1 and 2. The enclosed integrated
inspection report documents the inspection results which were discussed on January 7, and
February 12, 2009, with Mr. M. Skaggs 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 two NRC-identified findings and two self-revealing findings of very low
safety significance (Green). Three of the findings were determined to involve violations of NRC
requirements. Additionally, two licensee-identified violations, which were determined to be of
very low safety significance, are listed in this report. However, because of the very low safety
significance and because the violations were entered into your corrective action program, the
NRC is treating these violations as non-cited violations (NCVs) consistent with Section VI.A.1 of
the NRC Enforcement Policy. In addition, the NRC is exercising enforcement discretion in
accordance with Section VII.B.6, Violations Involving Special Circumstances, of the NRC
Enforcement Policy, and in accordance with Enforcement Guidance Memorandum 09-001, for a
violation of Technical Specification 3.4.15 involving the gaseous lower containment atmosphere
radioactivity monitor sensitivity. 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.
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/
Heather J. Gepford, Acting Chief
Reactor Projects Branch 6
Division of Reactor Projects
Docket Nos. 50-390, 50-391
License No. NPF-90 and Construction
Permit No. CPPR-92
Enclosure: NRC Inspection Report 05000390/2008005, 05000391/2008005,
05000390/2008501 w/Attachment: Supplemental Information
cc w/encl: (See page 3)
OFFICE
RII:DRP
RII:DRP
RII:DRP
RII:DRP
RII:DRS
RII:DRS
RII:DRS
SIGNATURE
JBB5
HJG1
RLM2
MEP2
CAP3
LEM
RRR1
NAME
JBaptist
HGepford
BMonk
MPribish
CPeabody
LMiller
RRodriguez
DATE
02/12/2009x
02/12/2009
02/12/2009
02/12/2009
02/12/2009
02/12/2009
02/12/2009
E-MAIL COPY?
YES
NO YES
NO YES
NO YES
NO YES
NO YES
NO YES
NO
OFFICE
RII:EICS
RII:DRS
RII:DRS
RII:
RII:
RII:DRS
RII:
SIGNATURE
SSparks for
RFA
BLC2
NAME
CEvans
RAiello
BCaballero
DATE
02/12/2009x
02/12/2009
02/12/2009
02/ /2009
02/ /2009
02/ /2009
02/ /2009
E-MAIL COPY?
YES
NO YES
NO YES
NO YES
NO YES
NO YES
NO YES
NO
3TVA
3
cc w/encl:
Gordon P. Arent
Manager
Watts Bar Unit 2
Watts Bar Nuclear Plant
Electronic Mail Distribution
Ashok S. Bhatnagar
Senior Vice President
Nuclear Generation Development and
Construction
Tennessee Valley Authority
Electronic Mail Distribution
Michael K. Brandon
Manager
Licensing and Industry Affairs
Tennessee Valley Authority
Electronic Mail Distribution
Preston D. Swafford.
Chief Nuclear Officer and Executive Vice
President
Tennessee Valley Authority
3R Lookout Place
1101 Market Street
Chattanooga, TN 37402-2801
Tom Coutu
Vice President
Nuclear Support
Tennessee Valley Authority
3R Lookout Place
1101 Market Street
Chattanooga, TN 37402-2801
General Counsel
Tennessee Valley Authority
Electronic Mail Distribution
John C. Fornicola
General Manager
Nuclear Assurance
Tennessee Valley Authority
Electronic Mail Distribution
Gregory A. Boerschig
Plant Manager
Watts Bar Nuclear Plant
Tennessee Valley Authority
Electronic Mail Distribution
Larry E. Nicholson
General Manager
Licensing & Performance Improvement
Tennessee Valley Authority
Electronic Mail Distribution
Michael A. Purcell
Senior Licensing Manager
Nuclear Power Group
Tennessee Valley Authority
Electronic Mail Distribution
Michael J. Lorek
Interim Vice President
Nuclear Engineering & Projects
Tennessee Valley Authority
Electronic Mail Distribution
Michael D. Skaggs
Site Vice President
Watts Bar Nuclear Plant
Tennessee Valley Authority
Electronic Mail Distribution
Mr. Fredrick C. Mashburn, Acting Manager
Corporate Nuclear Licensing and Industry
Affairs
Tennessee Valley Authority
4k Lookout Place
1101 Market Street
Chattanooga, Tennessee 37402-2801
Senior Resident Inspector
Watts Bar Nuclear Plant
U.S. Nuclear Regulatory Commission
1260 Nuclear Plant Road
Spring City, TN 37381-2000
County Executive
375 Church Street
Suite 215
Dayton, TN 37321
County Mayor
P.O. Box 156
Decatur, TN 37322
4
cc w/encl. (contd)
Lawrence Edward Nanney
Director
Division of Radiological Health
TN Dept. of Environment & Conservation
Electronic Mail Distribution
James H. Bassham
Director
Tennessee Emergency Management Agency
Electronic Mail Distribution
Ann Harris
341 Swing Loop
Rockwood, TN 37854
5
Letter to Preston D. Swafford from Heather J. Gepford dated February 12, 2009
SUBJECT:
WATTS BAR NUCLEAR PLANT - NRC INTEGRATED INSPECTION REPORT
05000390/2008005, 05000391/2008005, AND 05000390/2008501 AND
EXERCISE OF ENFORCEMENT DISCRETION
Distribution w/encl:
C. Evans, RII EICS
L. Slack, RII EICS
OE Mail
RIDSNRRDIRS
PUBLIC
P. Milano, NRR
Enclosure
U.S. NUCLEAR REGULATORY COMMISSION
REGION II
Docket Nos:
50-390, 50-391
License Nos:
NPF-90 and Construction Permit CPPR-92
Report Nos:
05000390/2008005, 05000391/2008005, 05000390/2008501
Licensee:
Tennessee Valley Authority (TVA)
Facility:
Watts Bar Nuclear Plant, Units 1 and 2
Location:
Spring City, TN 37381
Dates:
October 1, 2008 - December 31, 2008
Inspectors:
R. Monk, Senior Resident Inspector
C. Peabody, Acting Resident Inspector
M. Pribish, Resident Inspector
H. Gepford, Senior Health Physicist (Section 2PS1)
L. Miller, Senior Emergency Preparedness Inspector (Sections 1EP2,
1EP3, 1EP4, 1EP5, 4OA1, 4OA5.5)
R. Aiello, Senior Operations Engineer (Sections 1R11 and 4OA2.5)
B. Caballero, Operations Engineer (Sections 1R11 and 4OA2.5)
R. Rodriguez, Sr. Reactor Inspector (Section 4OA3.5)
Approved by:
Heather J. Gepford, Acting Chief
Reactor Projects Branch 6
Division of Reactor Projects
Enclosure
SUMMARY OF FINDINGS
IR 05000390/2008-005, 05000391/2008-005, 05000390/2008501; 10/01/2008 - 12/31/2008;
Watts Bar, Units 1 & 2; Plant Modifications, Identification and Resolution of Problems, and
Event Followup.
The report covered a three-month period of routine inspection by resident inspectors and
announced inspections by regional inspectors. Two NRC-identified Green findings and two self-
revealing Green findings were identified. Three of these findings were non-cited violations. The
significance of an issue is indicated by its color (Green, White, Yellow, Red) using the
Significance Determination Process in Inspection Manual Chapter (IMC) 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: Initiating Events
Green. The NRC identified a Green, non-cited violation of 10 CFR 50, Appendix
B, Criterion III, Design Control, for failure to translate revised design parameters
into the setpoint and scaling document for the lower containment particulate
radiation monitor. As a result, the radiation monitor was inoperable, due to
incorrect alarm setpoints, for longer than the Technical Specification allowed out
of service time. The licensee corrected the radiation monitor alarm setpoint and
initiated entered the issue into their corrective action program as Problem
Evaluation Report 154635.
The inspectors concluded that the finding was more than minor because the
radiation monitor inoperability resulted in potential impact on reactor safety and
adversely affected the availability and reliability of the equipment performance
attribute of the Initiating Events Cornerstone. This finding was evaluated using
the Significance Determination Process Phase 1 screening criteria and was
determined to be of very low safety significance because other methods of
reactor coolant system leak detection were available. The finding directly
involved the cross-cutting area of Problem Identification and Resolution under
the thorough evaluation of identified problems aspect of the corrective action
program component, in that, the licensee failed to properly evaluate the radiation
monitors as-found alarm setpoint, which was substantially different than the
specified setpoint, prior to resetting the alarm setpoint to the larger value (P.1.c).
(Section 4OA2.4)
Green. A Green self-revealing finding was identified for the failure to obtain
authorization prior to opening the main generator exciter field breaker
compartment and operating the de-latching bar. The licensees procedures for
controlling sensitive plant equipment specified that personnel obtain the Unit
Supervisors authorization prior to beginning work on sensitive equipment.
Operating the de-latching bar resulted in the exciter field breaker opening which
resulted in the turbine generator and the reactor tripping. The licensee entered
3
Enclosure
this issue into their corrective action program as Problem Evaluation Report
152955.
The finding was more than minor because it was associated with the Human
Performance attribute of the Initiating Events Cornerstone and adversely affected
the cornerstone objective to limit the likelihood of those events that upset plant
stability and challenge critical safety functions during at-power operations. This
finding was evaluated using the Significance Determination Process Phase 1
screening criteria and was determined to be of very low safety significance
because the finding did not contribute to both a reactor trip and the likelihood of
mitigation equipment or functions not being available. The cause of the finding
was directly related to the human performance and error prevention aspect of the
cross-cutting area of Human Performance, in that, personnel failed to use the
self-checking technique to stop and consider their actions for two minutes prior to
proceeding with an activity (H.4.a). (Section 4OA3.3)
Cornerstone: Mitigating Systems
Green. A Green self-revealing non-cited violation of 10 CFR 50 Appendix B,
Criterion III, Design Control, was identified for the failure to adequately translate
material specifications into procedures. As a result, the B-A essential raw cooling
water (ERCW) pump coupling failed due to an improper material being used.
The licensee entered this issue into their corrective action program as Problem
Evaluation Report 148716.
This finding is more than minor because it affects the plant modifications area of
the design control attribute of the Mitigating Systems Cornerstone objective of
reliability and availability, and if left uncorrected, it would result failure of other
ERCW pumps. This finding was evaluated using the Significance Determination
Phase 1 screening criteria and was determined to be of very low safety
significance because the finding did not represent an actual loss of safety
function of a single train of equipment for greater that its Technical Specification
allowed outage time. (Section 1R18.1)
Green. The NRC identified a Green, non-cited violation of Unit 1 Operating
License Condition 2.F for not having a carbon dioxide (CO2) suppression system
for the Unit 1 auxiliary instrumentation room with the capability to maintain the
design basis gas concentration of 50 percent in portions of the room for 15
minutes. The licensee entered the problem into their corrective action program.
The finding is more than minor because it affects the Mitigating Systems
cornerstone objective of ensuring reliability and capability of systems that
respond to initiating events and the cornerstone attribute of protection against
external factors, i.e. fire. The finding was determined to be of very low safety
significance by a Significance Determination Process Phase 1 evaluation. Test
records indicated a 50 percent CO2 concentration for 15 minutes in the lower half
of the room and a 45 percent concentration for 15 minutes at three quarters of
room height. This concentration was an acceptable amount to extinguish the
most likely fire in this portion of the room. (Section 4OA3.5)
4
Enclosure
B.
Licensee-Identified Violations
Two violations of very low safety significance, which were identified by the licensee,
were reviewed by the inspectors. Corrective actions taken or planned by the licensee
have been entered into the licensees corrective action program. The violations and
corrective action program tracking numbers are listed in Section 4OA7 of this report.
Enclosure
REPORT DETAILS
Summary of Plant Status
Unit 1 operated at or near 100 percent rated thermal power for the entire inspection period.
Restart of construction on Unit 2 began in December of 2007. Information on Watts Bar Unit 2
reactivation can be found at http://www.nrc.gov/reactors/plant-specific-items/watts-bar.html
1.
REACTOR SAFETY
Cornerstones: Initiating Events, Mitigating Systems, and Barrier Integrity
1R01 Adverse Weather Protection
.1
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 four 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. 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
b.
Findings
No findings of significance were identified.
.2
Readiness for Impending Adverse Weather Condition
a.
Inspection Scope
The inspectors reviewed the licensees preparation for and response to an actual freezing
condition on December 5, 2008. The inspectors verified performance and reviewed the
data associated with temperature monitoring of the RWST, which is required per licensee
procedure 1-PI-OPS-1-FP for outside air temperature less than 25 F. In addition, the
inspectors performed a walkdown of the RWST freeze protection enclosures to verify the
adequacy of construction and the operation of the installed temporary lighting.
6
Enclosure
b.
Findings
No findings of significance were identified.
1R04 Equipment Alignment
Partial 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 (OOS). The inspectors reviewed the functional
system descriptions, Updated Final Safety Analysis Report (UFSAR), system operating
procedures, and Technical Specifications (TSs) 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.
Partial walkdown of turbine-driven auxiliary feedwater (TDAFW) system following
component outage
Partial walkdown of 1B component cooling system (CCS) while the 1A CCS pump
was out OOS for motor preventive maintenance
Partial walkdown of the TDAFW pump while the 1A motor-driven auxiliary feedwater
pump was out of service for testing
b.
Findings
No findings of significance 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; SPP-10.10, Control of Transient Combustibles; and SPP-10.11,
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.
Cable spreading room
480 V reactor (RX) motor-operated valve (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 of significance were identified.
1R06 Flood Protection Measures
a.
Inspection Scope
The inspectors reviewed internal flood protection measures for the turbine building area.
Flooding in the turbine building could impact risk-significant components in the control
building if turbine building flood mitigation features were degraded. Turbine building flood
protection features were examined to verify that they were installed and maintained
consistent with the plant design basis. The inspectors reviewed the instrumentation and
associated alarms for turbine building floods to verify that the instrumentation was
periodically calibrated and that the respective alarms were appropriately integrated into
plant procedures. The inspectors also reviewed the licensee calculation for determining
maximum flood level in the turbine building for a condenser circulating water rupture and
licensee instructions for shutdown in the event of severe flooding to evaluate the availability
of structures, systems, or components (SSCs) for safe shutdown under worst case water
levels. Documents reviewed are listed in the Attachment to this report.
b.
Findings
No findings of significance were identified.
1R11 Licensed Operator Requalification
.1
Biennial Review
a.
Inspection Scope
The inspectors reviewed the facility operating history and associated documents in
preparation for this inspection. During the week of December 8, 2008, 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
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. The inspectors also
evaluated the licensees simulation facility for adequacy for use in operator licensing
examinations using ANSI/ANS-3.5-1985, American National Standard for Nuclear Power
Plant Simulators for use in Operator Training and Examination. The documentation
reviewed by the inspectors included written examinations, job performance measures
(JPMs), simulator scenarios, licensee procedures, on-shift records, simulator problem
report and performance test records, operator feedback records, licensed operator
qualification records, remediation plans, watch standing records, and medical records. The
records were inspected using the criteria listed in Inspection Procedure 71111.11.
Enclosure
8
Documents reviewed during the inspection are identified in the Attachment to this report.
The inspectors observed the licensee administer portions of the annual operating exam,
including three scenarios to one shift operating crew, and several JPMs. The inspectors
interviewed five licensed operators.
On December 19, 2008, the licensee completed the annual requalification operating tests
which are 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 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 of significance were identified.
.2
Resident Inspector Quarterly Review
a.
Inspection Scope
On October 21, 2008, the inspectors observed the simulator evaluation for scenario 3-OT-
SRT-E2-3B, Main Steam Line Leak/Break in Containment. The plant conditions led to a
Notice of Unusual Event emergency level classification.
The inspectors specifically evaluated the following attributes related to the crews
performance:
Clarity and formality of communication
Ability to take timely action to safely control the unit
Prioritization, interpretation, and verification of alarms
Correct use and implementation of abnormal operating instructions and emergency
operating instructions
Timely and appropriate emergency action level declarations per emergency plan
implementing procedures
Control board operation and manipulation including high-risk operator actions
Command and control provided by the unit supervisor and shift manager
The inspectors also attended the critique to assess the effectiveness of the licensee
evaluators and to verify that licensee-identified issues were comparable to issues identified
by the inspectors.
b.
Findings
No findings of significance were identified.
Enclosure
9
1R12 Maintenance Effectiveness
a.
Inspection Scope
The inspectors reviewed the two performance-based problems listed below. The focus of
the reviews was to assess the effectiveness of maintenance efforts that apply to 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).
125 V DC vital power a(1) performance improvement plan
(a)(1) classification of B-train auxiliary building gas treatment system
b.
Findings
No findings of significance were identified.
1R13 Maintenance Risk Assessments and Emergent Work Evaluation
a.
Inspection Scope
The inspectors evaluated, as appropriate, for the five 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; SPP-7.1, Work Control Process; and TI-124, Equipment to Plant Risk Matrix.
Emergency diesel generator (EDG) 2A-A maintenance while the A-A ERCW pump
was OOS
Planned maintenance on 1A CCS motor with A-A ERCW pump OOS
Emergent failure of A train main control room (MCR) air conditioner with B train
MCR air conditioner inoperable
Orange risk condition due to A-A ERCW pump unavailability combined with ERCW
pump coupling degradation and pressurizer power-operated relief valve (PORV) B
lock valve closed
Review of work week 607 activities with ERCW pump couplings degraded and one
PORV block valve closed
b.
Findings
No findings of significance were identified.
Enclosure
10
1R15 Operability Evaluations
a.
Inspection Scope
The inspectors reviewed five 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 LCOs and the risk significance in accordance with the SDP. The inspectors
verified that the operability evaluations were performed in accordance with SPP-3.1,
Corrective Action Program.
PER 148716, functional evaluation (FE) 42857, B-A ERCW pump shaft coupling
failure
PER 154635, Containment radiation monitor 1-RM-90-112 alarm setpoints found out
of tolerance
PERs 153738/153993, Incore instrument room containment penetration thermal
relief check valves 1-CKV-31-3907 and 3421 found stuck shut
PER 148716, FE 42961, ERCW pump continued operability with 410 SS couplings
installed
PER 154828, ERCW flow to spent fuel pump area cooler was less than TI-67.002
acceptance criteria
b.
Findings
No findings of significance were identified.
1R18 Plant Modifications
a.
Inspection Scope
The inspectors reviewed two permanent plant modifications to verify that design change
installation controls were adequate, affected operational procedures and licensing
documents were identified and revised accordingly, and that post-maintenance testing and
equipment return to service was adequate. Documents reviewed are listed in the
attachment.
Design change notice (DCN) S-10781-A and DCN S-08187-A, Revise ERCW pump
shaft material to XM-19 alloy
DCN 52631, Revise setpoint on lower containment radiation monitor gas channel
b.
Findings
.1
Essential Raw Cooling Water Pump Coupling
Introduction: A Green self-revealing non-cited violation of 10 CFR 50 Appendix B, Criterion
III, Design Control, was identified for the failure to adequately translate material
specifications into procedures which resulted in the failure of the B-A ERCW pump coupling
Enclosure
11
due to an improper material being used. The licensee entered this issue into their
corrective action program as PER 148716.
Description: On July 21, 2008, the B-A ERCW pump failed during operation due to the
shearing of a 410 stainless steel coupling caused by intergranular stress corrosion
cracking. The B-A ERCW pump was rebuilt, pre-service tested, and declared operable on
July 25, 2008.
The ERCW pumps were originally purchased with 410 stainless steel shafts and couplings.
Prior to plant operation, the licensee had issued DCN-S-10781-A to specify that the
preferred material for the ERCW pump shafts and couplings was XM-19 alloy. However,
procedure MI-67.001, Removal, Inspection and Repair of ERCW Pumps, which was revised
as a result of the design change process, lacked sufficient clarity to ensure that the
couplings would be replaced with XM-19 alloy. As a result, during the September 1995 B-A
ERCW pump overhaul, the shafts were replaced with XM-19 alloy but not the couplings.
Watts Bar Unit 1 began commercial operation in February 1996 with couplings of the
incorrect material installed on this pump, as well as, on other ERCW pumps.
Analysis: The licensees failure to adequately translate DCN S-10781-A material
specifications into the rebuild procedure was a performance deficiency, which resulted in
the failure of the B-A shaft coupling. This finding is more than minor because it affects the
plant modifications area of the design control attribute of the Mitigating Systems
Cornerstone objective of reliability and availability, and if left uncorrected, it would result in
the failure of other ERCW pumps. The inspectors evaluated this finding using IMC 0609,
Significance Determination Process, and determined that it was of very low safety
significance (Green) because the finding did not represent an actual loss of safety function
of a single train of equipment for greater that its TS allowed outage time. No cross-cutting
aspect was assigned because the cause of the finding was not indicative of current licensee
performance.
Enforcement: 10 CFR 50 Appendix B, Criterion III, Design Control, states, in part, that
design basis are correctly translated into specifications, drawings, procedures, and
instructions. Contrary to the above, the licensee failed to adequately translate design basis
into procedures, in that, material specifications for the ERCW pump couplings specified in
DCN S-10781-A were not properly incorporated into procedure MI-67.001. Because this
violation was of very low safety significance and it was entered into the licensees CAP as
PER 148716, this violation is being treated as an NCV, consistent with Section VI.A of the
NRC Enforcement Policy: NCV 5000390/2008005-01, Failure to Translate ERCW Pump
Coupling Material Change into Procedures.
.2
Technical Specification for the Containment Gaseous Radiation Monitors
Introduction: The inspectors identified a violation of TS 3.4.15, RCS Leakage Detection
Instrumentation, for the licensees failure to maintain the gaseous lower containment
atmosphere radioactivity monitor of the RCS leakage detection instrumentation operable.
The monitor had been inoperable since May 2000 as a result of not being able to perform
its safety function of detecting a reactor coolant pressure boundary leak of 1 gallon per
minute (gpm) in one hour due to improvements in reactor fuel quality. The NRC is
exercising enforcement discretion to not issue enforcement action for this violation in
accordance with Enforcement Guidance Memorandum (EGM) 09-001, Dispositioning
Enclosure
12
Violations of NRC Requirements for Operability of Gaseous Monitors for Reactor Coolant
System Leakage Detection.
Description: On October 31, 2008, the inspectors, after consultation with the Office of
Nuclear Reactor Regulation (NRR), informed the licensee that the gaseous lower
containment atmosphere radioactivity was not operable. The licensee initiated PER
155844, declared the equipment inoperable, complied with the applicable actions of TS 3.4.15 which allowed up to 30 days of continued operation with compensatory actions in
place, and submitted a license amendment request to change the TS. The TS amendment
was issued on November 25, 2008, which removed the requirement to maintain the
gaseous channel of the containment atmosphere radiation monitor as a method of RCS
leakage detection.
NRR determined that the technical bases for the gaseous lower containment atmosphere
radioactivity monitor to be operable included sufficient sensitivity to detect a reactor coolant
pressure boundary leak of 1 gpm in one hour. This sensitivity was consistent with the
information provided in NRC Information Notice (IN) 2005-024, Nonconservatism in Leak
Detection Sensitivity. This IN informed licensees that the 0.1-percent failed fuel
assumption (original source term for sensitivity calculations) introduced a nonconservatism
into the TS. However, the licensing bases for Watts Bar Unit 1 was not clear, in that, the
licensing basis documents acknowledged that, for fuel with little or no defects, this
sensitivity would not be expected. NRR considered that this circumstance would only
occur immediately after initial plant startup. However, the licensee mistakenly concluded
that the monitor would likewise be considered operable any time that fuel with little or no
defects was again in use, e. g., due to improved fuel quality.
In May 2000, the licensee developed calculation WBNTSR-062, Requirements for the
Containment Upper and Lower Compartment Radiation Monitors, which concluded that for
realistic RCS activity levels, the gaseous channel would not be capable of meeting the RG
1.45 detection sensitivity requirements. The UFSAR was revised to reflect this result and
the change was submitted to the NRC as part of its normal periodic update. This
conclusion was recently referenced in DCN 52631, dated June 20, 2008. In both cases, the
licensee failed to recognize that not meeting the required sensitivity resulted in the gaseous
lower containment atmosphere radioactivity monitors being inoperable. Contributing to the
failure to recognize this issue in June 2008 was the licensees mistaken perception that
since the NRC had been informed of the change by an UFSAR update, the change must
have been acceptable.
Analysis: The operation of Unit 1 in Modes 1-4 with one of the three required methods of
RCS leakage detection instrumentation required by TS 3.4.15 being inoperable was a
performance deficiency. The finding was more than minor because it was associated with
the Initiating Events Cornerstone attribute of equipment performance and affected the
cornerstone objective to limit the likelihood of those events that upset plant stability and
challenge critical safety functions during shutdown as well as power operations.
Specifically, the inoperability of a TS-required RCS leakage detection method affected the
likelihood of a loss of coolant accident initiator in keeping with the leak-before-break
concept. In EGM 09-001, the NRC states that the significance associated with a longer
response time (due to the lower sensitivity) is of very low safety significance. The EGM 09-
001 significance conclusion was based, in part, upon the availability of multiple and diverse
means for licensees to detect significant reactor coolant pressure boundary degradation
Enclosure
13
and take action to ensure continued public heath and safety. No cross-cutting aspect was
assigned.
Enforcement. TS 3.4.15 required, in part, that one lower containment atmosphere
radioactivity monitors (gaseous and particulate) be operable or restored to operable status
within 30 days, while in Modes 1, 2, 3, and 4. Contrary to this, between May 2000 and
November 25, 2008, the gaseous lower containment atmosphere radioactivity monitor was
inoperable while in Modes 1, 2, 3 and 4, in that, the containment atmosphere radioactivity
monitor was not capable of detecting a reactor coolant pressure boundary leak of 1 gpm in
one hour when radioactive gas content in the reactor coolant was low. Because this
violation was identified during the discretion period described in Enforcement Guidance
Memorandum 09-001, the NRC is exercising enforcement discretion in accordance with
Section VII.B.6, Violations Involving Special Circumstances, of the NRC Enforcement
Policy and is, therefore, not issuing enforcement action for this violation.
1R19 Post-Maintenance Testing
a.
Inspection Scope
The inspectors reviewed three post-maintenance test procedures, listed below, and/or test
activities, 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;
SPP-6.3, Pre-/Post-Maintenance Testing; and SPP-7.1, Work Control Process.
WO 08-819455-000, Replace D common station service transformer (CSST) auto
tap changer control relay
WO 08-823012-000, Repair of A train main control room chiller load control circuit
WO 08-816313-000, Card replacement on No. 2 vital battery charger
b.
Findings
No findings of significance were identified.
1R22 Surveillance Testing
a.
Inspection Scope
The inspectors witnessed four 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; SPP-8.2, 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.
Enclosure
14
Routine Surveillance Test:
WO 08-817807-000, 0-SI-82-12-B, Monthly diesel generator start and load test DG
WO 08-815234-000, 0-SI-215-41-A, Diesel generator 1A-A, 18-month service test
and battery charger test
In-Service Tests:
WO 08-822561-000, 1-SI-3-901-B, MDAFW (motor driven auxiliary feedwater) pump
B performance test
RCS leak detection
WO 08-817551-000, 1-SI-90-13, 92-day channel operational test of the containment
building lower compartment particulate radiation monitor loop 1-LPR-90-106A
b.
Findings
No findings of significance were identified.
Cornerstone: Emergency Preparedness
1EP2 Alert and Notification System Testing
a.
Inspection Scope
The inspector evaluated the adequacy of licensee=s methods for testing the alert and
notification system in accordance with NRC Inspection Procedure 71114, Attachment 02,
Alert and Notification System Evaluation. The applicable planning standard 10 CFR Part
50.47(b)(5) and its related 10 CFR Part 50, Appendix E, Section IV.D requirements were
used as reference criteria. The criteria contained in NUREG-0654, Criteria for Preparation
and Evaluation of Radiological Emergency Response Plans and Preparedness in Support
of Nuclear Power Plants, Revision 1, was also used as a reference.
The inspector reviewed various documents which are listed in the Attachment to this report.
This inspection activity satisfied one inspection sample for the alert and notification system
on a biennial basis.
b.
Findings
No findings of significance were identified.
1EP3 Emergency Response Organization Augmentation
a.
Inspection Scope
The inspector reviewed the licensee=s Emergency Response Organization (ERO)
augmentation staffing requirements and process for notifying the ERO to ensure the
readiness of key staff for responding to an event and timely facility activation. The
qualification records of key position ERO personnel were reviewed to ensure all ERO
qualifications were current. A sample of problems identified from augmentation drills or
Enclosure
15
system tests performed since the last inspection were reviewed to assess the effectiveness
of corrective actions.
The inspection was conducted in accordance with NRC Inspection Procedure 71114,
Attachment 03, Emergency Response Organization Staffing and Augmentation System.
The applicable planning standard, 10 CFR 50.47(b) (2) and its related 10 CFR 50,
Appendix E requirements were used as reference criteria.
The inspector reviewed various documents which are listed in the Attachment to this report.
This inspection activity satisfied one inspection sample for the ERO staffing and
augmentation system on a biennial basis.
b.
Findings
No findings of significance were identified.
1EP4 Emergency Action Level and Emergency Plan Changes
a.
Inspection Scope
Since the last NRC inspection of this program area, Revisions 87 and 88 of the Watts Bar
Emergency Plan were implemented based on the licensees determination, in accordance
with 10 CFR 50.54(q), that the changes resulted in no decrease in the effectiveness of the
Plan, and that the revised Plan continued to meet the requirements of 10 CFR 50.47(b) and
Appendix E to 10 CFR Part 50. The inspector conducted a sampling review of the Plan
changes and implementing procedure changes made between October 1, 2007 and
October 10, 2008 to evaluate for potential decreases in effectiveness of the Plan. However,
this review was not documented in a Safety Evaluation Report and does not constitute
formal NRC approval of the changes. Therefore, these changes remain subject to future
NRC inspection in their entirety.
The inspection was conducted in accordance with NRC Inspection Procedure 71114,
Attachment 04, Emergency Action Level and Emergency Plan Changes. The applicable
planning standard (PS), 10 CFR 50.47(b) (4) and its related 10 CFR 50, Appendix E
requirements were used as reference criteria.
The inspector reviewed various documents which are listed in the Attachment to this report.
This inspection activity satisfied one inspection sample for the emergency action level and
emergency plan changes on an annual basis.
b.
Findings
No findings of significance were identified.
1EP5 Correction of Emergency Preparedness Weaknesses and Deficiencies
a. Inspection Scope
The inspector reviewed the corrective actions identified through the Emergency
Preparedness program to determine the significance of the issues and to determine if
repeat problems were occurring. The facility=s self-assessments and audits were reviewed
Enclosure
16
to assess the licensee=s ability to be self-critical. In addition, the inspector reviewed
licensee self-assessments and audits to assess the completeness and effectiveness of all
emergency preparedness related corrective actions.
The inspection was conducted in accordance with NRC Inspection Procedure 71114,
Attachment 05, Correction of Emergency Preparedness Weaknesses. The applicable
planning standard, 10 CFR 50.47(b) (14) and its related 10 CFR 50, Appendix E
requirements were used as reference criteria.
The inspector reviewed various documents which are listed in the Attachment to this report.
This inspection activity satisfied one inspection sample for the correction of emergency
preparedness weaknesses on a biennial basis.
b.
Findings
No findings of significance were identified.
1EP6 Drill Evaluation
a.
Inspection Scope
The inspectors observed a licensee-evaluated emergency preparedness drill on November
5, 2008, involving a scenario that lead to a general emergency. The inspectors verified that
the emergency response organization was properly classifying the event in accordance with
Emergency Plan Implementing Procedure (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;
EPIP-4, Site Area Emergency; EPIP-5, General Emergency; and the Radiological
In addition, the inspectors verified that licensee evaluators were identifying deficiencies and
properly dispositioning performance against the performance indicator criteria in Nuclear
Energy Institute 99-02, Regulatory Assessment Performance Indicator Guideline.
b.
Findings
No findings of significance were identified.
2.
RADIATION SAFETY
Cornerstone: Public Radiation Safety (PS)
2PS1 Radioactive Gaseous and Liquid Effluent Treatment and Monitoring Systems
a.
Inspection Scope
Groundwater monitoring: The inspectors discussed current and future programs for
monitoring onsite groundwater with cognizant chemistry representatives including number
and placement of monitoring wells and identification of plant systems with the most
potential for contaminated leakage. The site has six onsite wells associated with the
radiological environmental monitoring program (REMP) and 37 non-REMP wells that are
Enclosure
17
used to monitor the onsite groundwater plume from two leaks identified in 2002. Recent
well sampling data and trends were evaluated. The inspectors reviewed and evaluated
procedural guidance for identifying and assessing onsite spills and leaks of contaminated
fluids. In addition, the inspectors reviewed the licensees 10 CFR Part 50.75(g) file and
compared the contents with known contaminated spill locations. The inspectors also
reviewed selected parts of the 2006 and 2007 Annual Radioactive Effluent Release Reports
with respect to abnormal releases or spills and releases with monitors OOS. Documents
reviewed are listed in the Attachment to this report.
b.
Findings
No findings of significance were identified.
4.
OTHER ACTIVITIES
4OA1 Performance Indicator Verification
a.
Inspection Scope
Cornerstone: Emergency Preparedness
The inspector sampled licensee submittals for three Performance Indicators (PI) listed
below. For each of the submittals reviewed, the inspector reviewed the period from
October 1, 2007 through June 30, 2008. To verify the accuracy of the PI data reported
during that period, PI definitions and guidance contained in Nuclear Energy Institute (NEI)
99-02, Regulatory Assessment Indicator Guideline, Revision 5, were used to verify the
basis in reporting for each data element.
Emergency Response Organization Drill/Exercise Performance (DEP)
Emergency Response Organization Readiness (ERO)
Alert and Notification System Reliability (ANS)
The inspectors reviewed portions of the raw PI data developed from monthly performance
indicator reports and discussed the methods for compiling and reporting the PIs with
cognizant emergency preparedness personnel. The inspector also independently screened
drill and exercise opportunity evaluations, drill participation reports, and drill evaluations.
Selected reported values were calculated to verify their accuracy. The inspectors
compared graphical representations from the most recent PI report to the raw data to verify
that the data was correctly reflected in the report. Reviewed documents are listed in the
Attachment to this report.
b.
Findings
No findings of significance were identified.
4OA2 Identification and Resolution of Problems
.1
Review of Items Entered into the Corrective Action Program
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
Enclosure
18
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 Inspection Procedure 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 2008 through December 2008,
although some examples expanded beyond those dates when the scope of the trend
warranted. Documents reviewed are listed in the Attachment to this report.
b.
Assessment and Observations
No findings of significance were identified. Two potential trends were identified from the
information reviewed.
The inspectors identified that the licensee missed including two PI entries into
quarterly reports to the NRC, one for an unplanned downpower and one for the high
head safety injection mitigating system performance index. The inspectors
reviewed the consequences surrounding each example and determined that in
neither case was the color of the PI affected and that both issues were
independently entered into the CAP as PERs 152109 and 152229.
The inspectors observed that the licensee had two instances of entering degraded
equipment into the CAP, without the recognition in the subsequent CAP process
that this equipment might require a TS or Offsite Dose Manual past operability
evaluation. Additionally, the inspectors identified that the licensees CAP process
review did not recognize a condition as potentially reportable, thus not obtaining an
Engineering review. In each case, the inspectors evaluated the PERs and
determined there were no significant consequences as the result of this problem..
.3
Annual Sample: Review of Operator Workarounds
a.
Inspection Scope
The inspectors reviewed the operator workaround program to verify that workarounds were
identified at an appropriate threshold, were entered into the CAP, and that corrective
actions were proposed or implemented. Specifically, the inspectors reviewed the licensees
workaround list and repair schedules, conducted tours, and interviewed operators about
required compensatory actions. Additionally, the inspectors looked for undocumented
workarounds, reviewed appropriate system health documents, and reviewed PERs related
to items on the workaround list.
Enclosure
19
b.
Findings and Observations
No findings of significance were identified.
.4
Annual Sample: Incorrect Alarm Setpoint for the Lower Containment Particulate Radiation
Monitor
a.
Inspection Scope
The inspectors reviewed the licensees assessment and corrective actions for PER 154635,
Incorrect Alarm Setpoint for the Lower Containment Particulate Radiation Monitor. The
PER was reviewed to ensure that the full extent of the issue was identified, an appropriate
evaluation was performed, and appropriate corrective actions were specified and prioritized.
b.
Findings
Introduction: The inspectors identified a Green, non-cited violation of 10 CFR 50, Appendix
B, Criterion III, Design Control, for failure to translate revised design parameters into the
setpoint and scaling document (SSD) for the lower containment particulate radiation
monitor. As a result, the radiation monitor was inoperable, due to incorrect alarm setpoints,
for longer than the Technical Specification allowed OOS time. The licensee corrected the
radiation monitor alarm setpoint and entered the issue into their corrective action program
as PER154635.
Discussion: The containment radiation monitors used for RCS leakage detection each
consist of a particulate and a gas detector channel. One monitor (1-RM-90-106) is normally
aligned to lower containment, while a redundant radiation monitor (1-RM-90-112) is aligned
to upper containment. TS 3.4.15, RCS Leakage Detection Instrumentation, only applies to
the radiation monitor which is aligned to lower containment.
On September 3, 2008, the 90-112 containment radiation monitor was aligned to lower
containment using preventive maintenance procedure PM 0639W, Conditional Calibration
To Be Performed If (90-112) Aligned To Lower Containment. PM 0639W is the procedure
used to ensure the correct alarm setpoints are selected for the particulate and gaseous
channels.
On October 14, 2008, during performance of licensee procedure 1-SI-90-19, 92 Day
Channel Operability Test Of Containment Building Upper Compartment Particulate
Radiation Monitor Loop 1-LPR-90-112A, the as-found alarm setpoint was out of tolerance.
The licensee initiated PER 154635 to document the discrepancy and adjusted the alarm
setpoint as specified in the radiation monitors SSD. PER 154635 description stated that
the as-found values were out of tolerance and the as-left values were in accordance with
the SSD; no specifics were given.
During a review of the completed 1-SI-90-19 paperwork, the inspectors found that the as-
found alarm setpoint for the particulate monitor was 1500 counts per minute (cpm) when
13,000 cpm was expected per the 90-112 radiation monitors SSD. The particulate
radiation monitor was returned to service with the alarm setpoint set at 13,000 cpm. The
inspectors determined that the 1500 cpm setpoint came from the 90-106 monitors SSD and
that PM 0639W had directed the use of 1500 cpm for the 90-112 monitors alarm setpoint.
Enclosure
20
On October 29, 2008, the licensee determined that containment radiation monitor alarm
setpoint changes made by design change 52631, dated July 11, 2008, had not been
properly incorporated into the SSD for the 90-112 particulate radiation monitor and, as a
result, the particulate radiation monitor would not be able to meet its TS specified safety
function of detecting a 1 gpm increase in RCS leakage in one hour if the SSD setpoint
(13,000 cpm) was selected. After the inspectors informed the licensee that the 90-112
particulate setpoint had been set at 13,000 cpm since October 14, 2008, the licensee
entered TS 3.4.15, RCS Leakage Detection Instrumentation.
Analysis: The inspectors determined that the continued operation of the containment
radiation monitor with an incorrect alarm setpoint was a performance deficiency. The
inspectors concluded that the finding was more than minor because the radiation monitor
inoperability resulted in a potential impact on reactor safety and adversely affected the
availability and reliability of the barrier integrity equipment performance attribute of the
Initiating Events Cornerstone.
The inspectors evaluated this finding using IMC 0609, Significance Determination Process,
and determined that it was of very low safety significance (Green) because other methods
of RCS leak detection were available and no actual leakage above 1 gpm was indicated
through the RCS water inventory surveillance. The finding directly involved the cross-
cutting area of Problem Identification and Resolution under the thorough evaluation of
identified problems aspect of the corrective action program component, in that, the licensee
failed to properly evaluate the radiation monitors as-found alarm setpoint of 1500 cpm,
which was substantially different than the 13,000 cpm specified setpoint, prior to resetting
the alarm setpoint to the larger value (P.1.c).
Enforcement: 10 CFR 50 Appendix B, Criterion III, Design Control, states, in part, that
measures shall be established to ensure that applicable regulatory requirements and the
design basis are correctly translated into specifications, drawings, procedures, and
instructions. Contrary to the above, the licensee failed to translate design basis information
into specifications, in that, the alarm setpoint, developed under DCN 52631, for the 90-112
containment radiation monitor was not incorporated in the SSD. Because this finding is of
very low safety significance and because it was entered into the licensees corrective action
program as PER 155844, this violation is being treated as an NCV, consistent with Section
VI.A of the NRC Enforcement Policy: NCV 05000390/2008005-02, Failure to Incorporate
Design Parameters into Plant Setpoint Document for the Containment Particulate Radiation
Monitor.
.5
Annual Sample: Biennial Licensed Operator Requalification Inspection
a.
Inspection Scope
The inspectors selected PERs 158392, 158697 and 158926 to verify that they correctly
described an issue related to excessive examination test item overlap and test item
duplication on remediation examinations for licensed operator written examinations.
b.
Findings and Observations
The inspectors conducted this review during the Biennial Licensed Operator Requalification
Inspection, which was conducted during the week of December 8, 2008. The licensee had
identified that 95% of one licensed operators retake examination was derived from a
Enclosure
21
previously administered exam of the current biennial written exam cycle. During an extent
of condition review, the licensee subsequently identified 3 additional operators who had
taken exams which excessively overlapped a previously administered exam. The
licensees procedure associated with licensed operator written examination development
was TRN-11.10, Annual Requalification Examination Development and Implementation,
Revision 13. Section 3.3, Step B of this procedure required that at least 50% of the
questions shall be different from the previous examinations developed for the same cycle.
The facility licensee initiated an apparent cause evaluation to determine why the decision
making process resulted in a departure from procedural requirements. The inspectors
verified that this issue had been accurately described in the licensees (CAP via PERs
158392, 158697 and 158926. The inspectors also verified that examination integrity
concerns were addressed by analyzing exam security agreements and operator test scores
with respect to test item duplication. However, the inspectors did not verify adequacy of the
corrective actions associated with the PERs because the licensee had not had sufficient
time to develop these actions.
4OA3 Event Followup
.1
(Closed) Licensee Event Report (LER) 05000390/2008-002-00, Manual Reactor Trip in
Response to Start of Feedwater Heater Isolation
On August 7, 2008, Unit 1 shut down to make a repair to the stator water cooling system.
With the plant near 53 percent power, operators secured both of the No. 7 heater drain tank
(HDT) pumps in accordance with the plant shutdown procedure. HDT level should have
been maintained by its bypass to condenser valve, but the level control valves air signal
line had failed so the valve failed to open. The high HDT level caused levels in the low
pressure feedwater heaters to rise until they reached the automatic isolation setpoint on all
three feedwater heater strings. The operators then manually tripped the plant. All systems
performed their intended safety functions in response to the trip. The LER was reviewed by
the inspectors, and no findings of significance were identified and no violation of NRC
requirements occurred. The licensee documented the failed equipment in PERs 149778
and 149790. This LER is closed.
.2
(Closed) LER 05000390/2008-003-00, Automatic Start of Auxiliary Feedwater Unavailable
during Startup Entry into Modes 2 and 1
On August 7, 2008, the NRC issued Integrated Inspection Report (IR) 05000390/2008003
which documented an NCV for inoperable auxiliary feedwater (AFW) automatic start
channels as required by TS 3.3.2, Function 6.e, start on trip of all main feedwater (MFW)
pumps. The finding was determined to be of very low safety significance because the
finding did not represent an actual loss of safety function of a single train for greater than its
TS-allowed outage time since other initiation signals were available to automatically start
the AFW pumps if needed. With this inspection report, NRC clarified that the
instrumentation channels must not only be capable of transmitting a trip signal but must
also reflect the actual operating condition of the MFW pumps.
The licensee documented this event in PER 147351. The enforcement aspect of this event
is documented in IR 05000390/2008003, Section 4OA2. This LER is closed.
Enclosure
22
.3
(Closed) LER 05000390/2008-004-00, Automatic Reactor Trip in Response to Opening of
Exciter Field Breaker
a.
Inspection Scope
On September 20, 2008, Unit 1 was tripped from 100% power due to a non-licensed
operator opening the Main Generator Exciter field breaker. The initial event follow-up was
conducted per inspection procedure 71153 and documented in Section 4OA3.2 of IR 05000390/2008004. The inspectors have subsequently reviewed the LER and associated
PER 152955, which included the root cause analysis and corrective action plans. The
inspectors also interviewed responsible Operations department personnel. Furthermore,
the inspectors verified that the corrective actions and extent of condition were consistent
with the root cause. This LER is considered closed.
b.
Findings
Introduction: A Green self-revealing finding was identified for the failure to obtain
authorization prior to opening the main generator exciter field breaker compartment and
operating the de-latching bar. Licensees procedures for controlling sensitive plant
equipment specified that personnel obtain the Unit Supervisors authorization prior to
beginning work on sensitive equipment. Operating the de-latching bar resulted in the
exciter field breaker opening which resulted in the turbine generator and the reactor
tripping. The licensee entered this issue into their corrective action program as PER
152955.
Description: On September 20, 2008, Unit 1 experienced a reactor trip from 100 percent
power. During rounds, a non-licensed operator stopped and opened the exciter field
breaker panel to show a trainee the breaker and explain that the breaker had to be
manually aligned and pushed into the cubicle while a second party pushed the de-latching
bar. The non-licensed operator pushed the de-latching bar during the explanation. This
resulted in the breaker opening, a turbine trip, and a reactor trip. All systems responded as
designed and performed their intended safety functions in response to the trip. The non-
licensed operator had not requested permission from control room personnel to either open
the breaker compartment door or operate the de-latching bar.
The licensee has established a self-imposed standard for controlling sensitive plant
equipment. TI-12.10, Control of Sensitive Equipment, states, in part, that the Shift Manager
or Unit Supervisor authorizes activities on sensitive equipment prior to work beginning. TI-
12.10 further states that the Exciter Field Breaker and controls, were components covered
by TI-12.10. The breaker compartment was labeled as Sensitive Equipment. Not obtaining
the Unit Supervisors permission before opening the main generator exciter field breaker
compartment and operating the breaker de-latching bar, as specified in TI-12.10, was
considered a Finding.
Analysis: Operating the de-latching bar on the exciter field breaker while the breaker was
closed was a performance deficiency which resulted in a reactor scram. The finding was
more than minor because it was associated with the human performance attribute of the
Initiating Events Cornerstone and adversely affected the cornerstone objective to limit the
likelihood of those events that upset plant stability and challenge critical safety functions
during at-power operations. This finding was evaluated using the SDP Phase 1 screening
criteria and was determined to be a finding of very low safety significance (Green) because
Enclosure
23
the finding did not contribute to both a reactor trip and the likelihood of mitigation equipment
or functions would not be available.
The cause of the finding was directly related to the human performance and error
prevention aspect of the cross-cutting area of Human Performance, in that, personnel failed
to use a self-checking technique, the two minute rule. The two minute rule required
personnel to stop and consider their actions for two minutes prior to proceeding with an
activity (H.4(a)).
Enforcement: Enforcement action does not apply because the performance deficiency did
not involve a violation of a regulatory requirement. Because this finding does not involve a
violation of regulatory requirements and has very low safety significance, it is identified as
FIN 05000390/2008005-03, Performing Non-Authorized Activities on Exciter Field Breaker
Results In Reactor Trip.
.4
(Closed) LER 05000390/2008-005-00, Report of Inoperability of Radiation Monitor due to
Non-conservative Setpoint
On October 29, 2008, a discrepancy in the setpoint was identified for the particulate
channel of the radiation monitor being credited for meeting TS 3.4.15, Leakage Detection
Instrumentation. From October 14 to October 29, 2008, the RCS leakage detection system
had been inoperable due to this incorrect setpoint. Consequently, the licensee had been
operating in a condition prohibited by technical specifications. The enforcement aspects of
violation are discussed in section 4OA2.4 of this report. This LER is closed.
.5
(Closed) Unresolved Item (URI)05000390/2007007-01, Carbon Dioxide System in FA 48
Appears to Deviate From Design Criterion in SSER
Introduction: The inspectors identified a Green NCV of Unit 1 Operating License Condition
2.F for the failure of the installed carbon dioxide (CO2) fire suppression system to deliver
and maintain the design basis gas concentration of 50 percent for 15 minutes in portions of
Fire Area (FA) 48, the auxiliary instrumentation room.
Description: Alternative shutdown was selected for the auxiliary instrument room and thus
a CO2 gas suppression system was installed to meet the requirements of 10 CFR 50
Appendix R,Section III.G.3. The CO2 system was required to be designed in accordance
with National Fire Protection Association standard-12 (NFPA 12), Standard on Carbon
Dioxide Extinguishing Systems; and the Watts Bar Fire Protection Report (FPR) as
approved in NRC Supplemental Safety Evaluation Report (SSER) No. 18 (NUREG 0847).
The applicable edition of NFPA-12 specified 50 percent concentration for deep-seated fires,
but did not specify a definite hold time. The approved SSER stated that the CO2 system
must achieve a concentration of at least 50 percent within seven minutes of initiation and
hold that concentration for 15 minutes. The concentration values must be achieved at any
point in the room where combustibles capable of deep seated fires are located. The basis
for these values in the SSER was testing performed by Sandia National Laboratory on deep
seated fires and CO2 systems as described in NRC IN 92-28, Inadequate Fire Suppression
System Testing, issued April 8, 1992. The inspectors reviewed records of a discharge test
conducted at the time of initial CO2 system installation. The test records indicated that 50
percent CO2 concentration for 15 minutes was achieved in the lower half of the room and
45 percent concentration was held for 15 minutes at the three quarters of room height level.
Enclosure
24
Therefore, the approved SSER concentration for the upper portion of the room was not met.
The licensee initiated PER 125632 to address this issue.
Analysis: The finding was a performance deficiency because the licensee failed to meet
their NFPA code of record and it was within their ability to identify and correct. The finding
was more than minor because it was associated with the reactor safety Mitigating Systems
cornerstone attribute of protection against external factors, i.e. fire, and it affected the
objective of ensuring reliability and capability of systems that respond to initiating events.
The finding was screened as of very low safety significance by a SDP Phase 1 evaluation,
in accordance with IMC 0609, Appendix F, Fire Protection Significance Determination
Process. This was due to assigning a low degradation factor to the deficient CO2 system
based upon:
Test records indicated a 50 percent CO2 concentration in the lower half and 45
percent at three quarters of the rooms height. These concentrations lasted for 15
minutes.
Most of the ignition sources were located in the lower portion of the room where the
required concentration was maintained.
Fire spread from the lower portion of the room was difficult since a great majority of
the targets (electrical cables) were located in conduits or enclosed raceways.
These configurations preclude a secondary ignition source outside of the 50
percent CO2 concentration zone.
The only ignition sources in the upper portion of the room were thermoplastic
cables capable of self-igniting. However, their failure would not affect the credited
SSD strategy of alternative shutdown which used equipment powered by other
cables not located in this room.
Since the CO2 concentration issue occurred during the original installation of the CO2
suppression system, the issue was not indicative of licensee current performance and no
cross-cutting aspect was assigned.
Enforcement: Watts Bar Unit 1 License Condition 2F requires that the licensee implement
and maintain in effect all provisions of the approved fire protection program, as approved in
Supplements 18 and 19 of the SER (NUREG-0847). These documents incorporate the
requirements of 10 CFR 50, Appendix R, Section III.G.3. This section of Appendix R
requires a fixed fire suppression system for the auxiliary instrumentation room area since it
contains safe shutdown equipment and alternative safe shutdown was selected for this
area. The Watts Bar CO2 gas suppression system was required to be designed in
accordance with NFPA 12, 1973 Edition and the SSER No.18. NFPA 12, 1973, specified
that an acceptable CO2 system deliver and hold a minimum gas concentration of 50 percent
and the SSER stated that this concentration must be held for 15 minutes.
Contrary to the above, since receipt of the operating license on February 7, 1996, until the
present, the CO2 system for the auxiliary instrumentation room was not designed in
accordance with the 1973 Edition of NFPA 12 and SSER No. 18, in that, the CO2 system
was unable to deliver and maintain a minimum gas concentration of 50 percent in the upper
portion of the room for 15 minutes. Because this finding is of very low safety significance
Enclosure
25
and has been entered into the licensees corrective action program as PER 125632, this
finding is being treated as an NCV, consistent with Section VI.A.1 of the NRCs
Enforcement Policy: NCV 05000390/2008005-04, Carbon Dioxide System in Fire Area 48
Failed to Meet Design Criterion.
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 reviews and inspection activities.
b.
Findings
No findings of significance were identified.
.2
(Closed) NRC Temporary Instruction (TI) 2525/175, Emergency Response Organization,
Drill/Exercise Performance Indicator, Program Review
The inspectors completed TI 2515/175. Appropriate documentation of the results was
provided to NRC, HQ, as required by the TI. This completed the Region II inspection
requirements of this TI for the Watts Bar Nuclear Plant.
4OA6 Meetings, including Exit
The inspectors presented the inspection results to Mr. M. Skaggs and other members of
licensee management at the conclusion of the inspection on January 7, and again on
February 12, 2009. The inspectors asked the licensee whether any materials examined
during the inspection should be considered proprietary. No proprietary information was
identified.
4OA7 Licensee-Identified Violations
The following violations of very low safety significance (Green) were identified by the
licensee and are violations of NRC requirements which meet the criteria of Section VI of the
NRC Enforcement Policy, NUREG-1600, for being dispositioned as NCVs.
TS 5.2.2.3, Administrative Controls Section, required that the Operations
Superintendent shall have a valid senior reactor operators (SRO) license. During
the time period of March 26, 2008, until October 17, 2008, the Operations
Superintendent had an expired SRO license. This finding is of very low safety
significance because the Operations Superintendent attended all required training
Enclosure
26
and completed successfully all required examinations during the expired period.
This issue was entered in the licensees CAP as PER 155152.
10 CFR 55.25 states If, during the term of the license, the licensee develops a
permanent physical or mental condition that causes the licensee to fail to meet the
requirements of § 55.21 of this part, the facility licensee shall notify the Commission,
within 30 days of learning of the diagnosis, in accordance with § 50.74(c). Contrary
to the above, on October 21, 2008, the licensee discovered they had failed to notify
the Commission within 30 days after one licensed operator had a permanent
change in physical medical condition, as required by 10 CFR 55.25. This finding
was evaluated using the traditional enforcement process because it impacted the
Commissions ability to perform its regulatory licensing function. This finding was of
very low safety significance because the medical condition was under control and
had no impact on the individuals ability to perform licensed duties. The licensee
entered this issue into their CAP as PERs 155159 and 155130.
ATTACHEMENT: SUPPLEMENTAL INFORMATION
Attachment
SUPPLEMENTAL INFORMATION
KEY POINTS OF CONTACT
Licensee personnel
L. Belvin, Radiation Protection Manager
G. Boerschig, Plant Manager
M. Brandon, Licensing and Industry Affairs Manager
R. Crews, Operations Training Manager
T. Coutu, Vice President, Nuclear Support
T. Detchemedy, Emergency Preparedness Manager
N. Good, Simulator Services Supervisor
B. Hunt, Operations Superintendent
B. Marks, Corporate Emergency Preparedness Manager
G. Mauldin, Site Engineering Manager
M. McFadden, Site Nuclear Assurance Manager
T. Newman, Operations Training Contractor
S. Reininghaus, Operations Training Contractor
A. Scales, Operations Manager
M. Skaggs, Site Vice President
W. Thompson, Training Manager
J. Tortura, Site Support
D. Voeller, Maintenance and Modifications Manager
ITEMS OPENED, CLOSED, AND DISCUSSED
Opened and Closed 05000390/2008005-01
Failure to Translate ERCW Pump Coupling Material
Change into Procedures (Section 1R18.1)05000390/2008005-02
Failure to Incorporate Design Parameters into Plant
Setpoint Document for the Containment Particulate
Radiation Monitor (Section 4OA2.4)05000390/2008005-03
Performing Non-Authorized Activities on Exciter Field
Breaker Results In Reactor Trip (Section 4OA3.3)05000390/2008005-04
Carbon Dioxide System in Fire Area 48 Failed to Meet
Design Criterion (Section 4OA3.5)
Closed
05000390/2008-002-00
LER
Manual Reactor Trip in Response to Start of
Feedwater Heater Isolation (Section 4OA3.1)
2
Attachment
05000390/2008-003-00
LER
Automatic Start of Auxiliary Feedwater Unavailable
During Startup Entry into Modes 2 and 1 (Section
4OA3.2)
05000390/2008-004-00
LER
Automatic Reactor Trip in Response to Opening of
Exciter Field Breaker (Section 4OA3.3)
05000390/2008-005-00
LER
Report of Inoperability of Radiation Monitor due to
Non-conservative Setpoint (Section 4OA3.4)05000390/2007007-01
Carbon Dioxide System in FA 48 Appears to
Deviate From Design Criterion in SSER (Section
4OA3.5)
2515/175
TI
Emergency Response Organization, Drill/Exercise
Performance Indicator, Program Review (Section
4OA5.2)
Discussed
None
LIST OF DOCUMENTS REVIEWED
Section 1R01: Adverse Weather Protection
PER 158406 - Freeze protection discrepancies for the month of November.
PER 156045 - Freeze protection for RWST not checked
WO 08-812230-000
Section 1R06: Flood Protection Measures
Watts Bar Unit 1 Individual Plant Examination, Appendix E, Section 1.4.3, Turbine Building (flood
analysis)
PMUG 2127F, Functional Check and Calibration of Flood Mode Switches (1-LS-040-0019, Unit 1
Condenser Pit Flood Detector)
Design Criteria WB-DC-40-29, Flood Protection Provisions
Annunciator Response Instruction, ARI-166-172, Miscellaneous & HPFP, Page 13 of 48, response
for TURB/AUX/RX BLDG FLOODED.
Calculation WBNAPS2-165, Turbine Building Flooding Due To A Break In The Condenser
Vendor Technical Document WBN-VTD-D925-0090, Mercoid Liquid Level Control Switches
Section 1R11: Licensed Operator Requalification
Procedures:
TRN-12, Simulator Regulatory Requirements, Rev. 8
TRN-11.10, Annual Examination Development and Administration, Rev. 13
TRN-11.4.Continuing Training for Licensed Individuals, Rev. 14
TRN-11.11 Requalification Periodic Written Exam Development & Implementation, Rev. 6
TRN-11.7 Simulator Exercise Guide Development and Revision, Rev. 3
Written Examinations Reviewed:
All 2007 Biennial Written Examinations
3
Attachment
Simulator Documents:
TVA Simulator Services Group Directive, Core Model Evaluations, 11/19/08
TVA Simulator Services Group Directive, Simulator Testing Program, 06/17/08
Closed Simulator Problem Reports since 2006
Outstanding Simulator Problem Report List as of 01/01/2008
Transient Tests
Transient No. 8: Max Sized LOCA w/ LOOP (2007 & 2008)
Transient No. 10: RCS Depressurization to Saturation Using PORV w/o HP ECCS (2007 & 2008)
Malfunction Tests:
FW23, Main Feedwater Break Inside Containment (2002 & 2006)
MS02, Main Steam Line Break Outside of Containment (2002 & 2006)
RH01, RHR Pump Trip or Fails (2002 & 2006)
TU02, Main Turbine High Vibration (2002)
TC09, Main Turbine Trip on Low Bearing Oil Pressure (2006)
TH05, Steam Generator Tube Failure (2004 & 2008)
Normal Evolutions Tests:
Cycle 8 Core Reload Test Packages (9)
Job Performance Measures (JPMs)
3-OT-JPMAADMIN1, Demonstrate Knowledge of Admin/Rad Procedure, Rev. 6
3-OT-JPMA015, Local Operation of Turbine Driven AFW Pump, Rev. 7
3-OT-JPMA001B, Local Restart of Control & Service Air Compressors, Rev. 4
3-OT-JPMR069A, Transfer ECCS to RHR Containment Sump, Rev. 2
3-OT-JPMS082A, Classify the Event (Loss of Annunciators), Rev. 7
3-OT-JPMR039, Start Thermal Barrier Booster Pump, Rev. 9
3-OT-JPMR168, Respond to Multiple Dropped Rod, Rev. 3
Simulator Scenarios
3-OT-SRE0019, Steam Generator Tube Rupture, Rev. 9
3-OT-SRE0020, MSL Break I/S Containment w/ Loss of Containment Spray, Rev. 6
3-OT-SRE0006B, ATWS/Stm Line Break (O/S Containment) Loss of Offsite Power, Rev. 4
Problem Evaluation Reports (PERs)
PER 117527, Core Model Impact on I Limits During Simulator Scenarios
PER 139711, Simulator RVLIS & Rod Step Counters
PER 144939, Steam Generator Tube Rupture Pressure Response on Simulator
PER 138223, Questions Regarding Simulator Steam Generator Tube Rupture Response
PER 148319, WBN Simulator Out of Service
PER 155130, Expired SRO License
PER 155159, SRO Change in Medical Condition
PER 158392, Retake Biennial Exam w/ Excessive Overlap & Test Item Duplication
PER 158697, Additional Biennial Exams w/ Excessive Overlap
PER 158926, Test Item Duplication on Weekly LOR Retake Exams
Other:
Attendance Records (4)
Reactivation Records (4)
Medical Records (10)
4
Attachment
Feedback Comments from Licensed Operator Requal 2006 thru 2008
Remedial Packages (4)
Section 1R18: Plant Modifications
DCN-50107, Revise High Radiation Alarm Setpoint to Allow for Changes in Background Radiation
DCN-52631, Revise Setpoint on Gas Channel
WBNTSR-062, Requirements for the Containment Upper and Lower Compartment Radiation
Monitors
Section 1EP2: Alert and Notification System Testing
Procedures and Documentation
EPFS-9, Inspection, Service, and Maintenance of the Prompt Notification System (PNS) at Browns
Ferry, Sequoyah, and Watts Bar Nuclear Plants, Rev. 2
EPIL-18, Evaluation of Changes to Alert and Notification Systems (ANS), Rev. 1
Records and Data
PNS Checklist and Trouble Reports, September 06, 2006 - October 1, 2008
Annual Maintenance documentation, April 1, 2007 - June 3, 2007
Section 1EP3: Emergency Response Organization Augmentation
Procedures
EPIL-14, Facilitation of the Alert & Notification System and Pager Tests, Rev. 13
Records and Data
January 9, 2007 to October 14, 2008, Weekly Emergency Paging Systems Tests
March 15, 2007, REP Drill - Blue Team
May 10, 2007, REP Drill - Red Team
March 31, 2008, REP Drill - Green Team
May 29, 2008, REP Drill - Orange Team
September 15, 2008, REP Drill - Blue Team
September 11, 2008, Annual Emergency Preparedness Medical Drill - Rhea County Medical
Center and Rhea County Emergency Medical Service
September 18, 2007, Annual Emergency Preparedness Medical Drill - Rhea County Medical
Center and Rhea County Emergency Medical Service
December 11, 2007, Annual Emergency Preparedness Medical Drill - Athens Regional Hospital
Section 1EP4: Emergency Action Level and Emergency Plan Changes
Tennessee Valley Authority Nuclear Radiological Emergency Plan, Rev. 87 and 88
EPIL-1, Procedures, Maps and Drawings, Rev. 25
Plans and Changes packages
EPIP-1, Emergency Plan Classification Flowchart, Rev. 28 and 29
5
Attachment
Section 1EP5: Correction of Emergency Preparedness Weaknesses and Deficiencies
Audits and Self-Assessments
NA-CH-07-003, Assessment of Emergency Preparedness Performance, June 2007
SSA0804, Radiological Emergency Preparedness Program Audit Report, May 19 - August 22,
2008
WBN-SIT-08-013, Emergency Preparedness Program Self-Assessment Report, April 28 - May 2,
2008
WBN-SIT-08-020, B5b Phase 2 and 3 Implementation Self-Assessment Report, April 28 - May 2,
2008
WBN-SIT-08-015, Emergency Equipment Inventories Self-Assessment Report, December 17 - 20,
2008
PER Summary of Corrective Actions
128350, 130252,130346, 130383, 130385, 130388, 130457, 133561, 133625, 134131, 136400,
137996, 138725, 141449, 142644, 145306, 145742, 155274,155275, 155276, 155277, 155373,
155374, 155376, 155377, 155414
Section 2PS1: Radioactive Gaseous and Liquid Effluent Treatment and Monitoring Systems
Procedures, Guidance Documents, and Reports
2006 Annual Radioactive Effluent Release Report
2007 Annual Radioactive Effluent Release Report
2006 Annual Environmental Radiological Operating Report
2007 Annual Environmental Radiological Operating Report
Offsite Dose Calculation Manual, Rev. 20
SPP-5.14, Guide for Communicating Inadvertent Radiological Spills/Leaks to Outside Agencies,
Rev. 1
RCDP-11, Protocol for Remediation of Inadvertent Spills or Leaks of Contaminated Liquids, Rev. 0
0-PI-CEM-11.0, Monitoring Well Sampling, Rev. 1
Arcadis Presentation, 4/16/04
Records and Data
Groundwater monitoring well results, calendar years 2007 and 2008
Corrective Action Program Documents
PER 146594, Well tritium above acceptance criteria, 6/11/08
PER 134706, Well K tritium, 12/04/07
PER 125208, Well K and L above acceptance criteria, 5/22/07
Section 4OA1: Performance Indicator Verification
Procedures
EPIL-15, Emergency Preparedness Performance Indicators, Rev. 12
Records and Data
DEP data from 4th Qtr 2007 to 2nd Qtr 2008
6
Attachment
ERO data from 4th Qtr 2007 to 2nd Qtr 2008
ANS data from 4th Qtr 2007 to 2nd Qtr 2008
Section 4OA2: Identification and Resolution of Problems
NADP-3, Managing the Operating Experience Program
PERs written as a result of NRC identified issues
149257
Water is leaking from the ceiling near 2B-1B RX MOV board
151046
No protected equipment sign on the door of the 1A SIP room, although the 1B SIP
was OOS and 1A was protected
148243
Inadequate instructions for replacing and installation of the controller and no PMT
specified for Primary Water Blender Flow Control
151026
AUO did not have keys to spaces required for EOP actions in a timely manner
151252
CAP process failed to recognize an operability issue (ODCM TS 2.0.3 not met) when
Steam Generator Blowdown effluent release valve (1-FCV 15-44) was found out of
surveillance grace
151962
PER Screening Committee composition not IAW PIDP-4
152038
Insulation missing from 1B RHR Hx
152109
Missed Unplanned Transient input
152229
Missed MSPI failure input of 1B CCP
152372
CAP process failed to recognize an operability issue when RM-106 found inop
during calibration
153779
LCO time tracking error
155193
Conduit separation inadequacy
155524
PER screening committee failed to ensure secondary boundary doors left ajar was
not reportable
155844
DCN output inadequately captured in implementing procedure
155046
Failure to comply w/ Tech Specs for RCS leak detection gaseous monitors
156371
PER Screening Committee untimely in processing request for Functional Evaluation
159025
B EBR chiller TCV indicated fully open w/ chiller not running
159474
Scaffold w/ insufficient (0) clearance to safety related equipment
159743
Vendor drawings and DCAs do not reflect the as built configuration of ERCW pump
Shaft Coupling materials.
159751
Lack of timeliness of communication of potential issues found on Unit 2 that could
affect Unit 1
Section 4OA5: Other Activities
PER 125632, NRC SER difference with docket
Fire Protection Program Change Regulatory Review for PER 125632