ML071640143
| ML071640143 | |
| Person / Time | |
|---|---|
| Site: | Braidwood  |
| Issue date: | 06/12/2007 |
| From: | Richard Skokowski NRC/RGN-III/DRP/RPB3 |
| To: | Crane C Exelon Generation Co, Exelon Nuclear |
| References | |
| FOIA/PA-2010-0209 IR-07-007 | |
| Download: ML071640143 (36) | |
See also: IR 05000456/2007007
Text
June 12, 2007
Mr. Christopher M. Crane
President and Chief Nuclear Officer
Exelon Nuclear
Exelon Generation Company, LLC
4300 Winfield Road
Warrenville, IL 60555
SUBJECT: BRAIDWOOD STATION, UNITS 1 AND 2
NRC PROBLEM IDENTIFICATION AND RESOLUTION
INSPECTION REPORT 05000456/2007007 AND 05000457/2007007
Dear Mr. Crane:
On May 1, 2007, the U.S. Nuclear Regulatory Commission (NRC) completed a team inspection
of problem identification and resolution at your Braidwood Station, Units 1 and 2. The enclosed
inspection report documents the inspection findings which were discussed on March 30, 2007,
Mr. M. Smith and other members of your staff, and subsequently on May 1, 2007, with
Mr. D. Ambler and other members of your staff.
This inspection was an examination of activities conducted under your license as they
relate to the identification and resolution of problems, compliance with the Commissions
rules and regulations, and with the conditions of your operating license. Within these areas,
the inspection involved selected examination of procedures and representative records,
observations of activities, and interviews with personnel.
On the basis of the sample selected for review, the team concluded that, in general, problems
were properly identified, evaluated, and corrected. However, the inspectors identified two
findings during the inspection. One finding of very low safety significance (Green) was
identified for the licensees failure to perform an adequate extent of condition review for safety
related valves that had not been included in and tested in accordance with the inservice test
program. The second finding involved the licensees failure to maintain an adequate operations
procedure that had the potential to secure the only remaining residual heat removal pump while
in the recirculation mode of operation. Both findings were violations of NRC requirements.
However, because each finding was of very low safety significance and because the findings
were entered into your corrective action program, the NRC is treating these findings as non-
cited violations (NCVs), in accordance with Section VI.A.1 of the NRCs Enforcement Policy.
C. Crane -2-
If you contest the subject or severity of a non-cited violation, you should provide a
response within 30 days of the date of this inspection report, with the basis for your
denial, to the U.S. Nuclear Regulatory Commission, ATTN: Document Control Desk,
Washington, DC 20555-0001; with a copy to the Regional Administrator, U.S. Nuclear
Regulatory Commission - Region III, 2443 Warrenville Road, Suite 210, Lisle, IL
60532-4352; the Director, Office of Enforcement, U.S. Nuclear Regulatory Commission,
Washington, DC 20555-0001; and the Resident Inspector Office at the Braidwood Station.
In accordance with 10 CFR 2.390 of the NRC's "Rules of Practice," a copy of this letter and
its enclosure will be available electronically for public inspection in the NRC Public Document
Room or from the Publicly Available Records (PARS) component of NRC's document system
(ADAMS), accessible from the NRC Web site at http://www.nrc.gov/reading-rm/adams.html
(the Public Electronic Reading Room).
Sincerely,
/RA/
Richard A. Skokowski, Chief
Branch 3
Division of Reactor Projects
Docket Nos. 50-456; 50-457
Enclosure: Inspection Report No. 05000456/2007007 and 05000457/2007007
w/Attachment: Supplemental Information
cc w/encl: Site Vice President - Braidwood Station
Plant Manager - Braidwood Station
Regulatory Assurance Manager - Braidwood Station
Chief Operating Officer
Senior Vice President - Nuclear Services
Vice President - Operations Support
Vice President - Licensing and Regulatory Affairs
Director Licensing
Manager Licensing - Braidwood and Byron
Senior Counsel, Nuclear, Mid-West Regional
Operating Group
Document Control Desk - Licensing
Assistant Attorney General
Illinois Emergency Management Agency
State Liaison Officer
Chairman, Illinois Commerce Commission
C. Crane -2-
If you contest the subject or severity of a non-cited violation, you should provide a
response within 30 days of the date of this inspection report, with the basis for your
denial, to the U.S. Nuclear Regulatory Commission, ATTN: Document Control Desk,
Washington, DC 20555-0001; with a copy to the Regional Administrator, U.S. Nuclear
Regulatory Commission - Region III, 2443 Warrenville Road, Suite 210, Lisle, IL
60532-4352; the Director, Office of Enforcement, U.S. Nuclear Regulatory Commission,
Washington, DC 20555-0001; and the Resident Inspector Office at the Braidwood Station.
In accordance with 10 CFR 2.390 of the NRC's "Rules of Practice," a copy of this letter and
its enclosure will be available electronically for public inspection in the NRC Public Document
Room or from the Publicly Available Records (PARS) component of NRC's document system
(ADAMS), accessible from the NRC Web site at http://www.nrc.gov/reading-rm/adams.html
(the Public Electronic Reading Room).
Sincerely,
Richard A. Skokowski, Chief
Branch 3
Division of Reactor Projects
Docket Nos. 50-456; 50-457
Enclosure: Inspection Report No. 05000456/2007007 and 05000457/2007007
w/Attachment: Supplemental Information
cc w/encl: Site Vice President - Braidwood Station
Plant Manager - Braidwood Station
Regulatory Assurance Manager - Braidwood Station
Chief Operating Officer
Senior Vice President - Nuclear Services
Vice President - Operations Support
Vice President - Licensing and Regulatory Affairs
Director Licensing
Manager Licensing - Braidwood and Byron
Senior Counsel, Nuclear, Mid-West Regional
Operating Group
Document Control Desk - Licensing
Assistant Attorney General
Illinois Emergency Management Agency
State Liaison Officer
Chairman, Illinois Commerce Commission
DOCUMENT NAME: C:\FileNet\ML071640143.wpd
G Publicly Available G Non-Publicly Available G Sensitive G Non-Sensitive
To receive a copy of this document, indicate in the concurrence box "C" = Copy without attach/encl "E" = Copy with attach/encl "N" = No copy
OFFICE RIII RIII
NAME DSmith:dtp RSkokowski
DATE 06/12/2007 06/12/2007
OFFICIAL RECORD COPY
Letter to Christopher M. Crane from Richard A. Skokowski dated June 12, 2007
SUBJECT: BRAIDWOOD STATION, UNITS 1 AND 2
NRC PROBLEM IDENTIFICATION AND RESOLUTION
INSPECTION REPORT 05000456/2007007; 05000457/2007007
DISTRIBUTION:
RAG1
TEB
RFK
RidsNrrDirsIrib
GEG
KGO
RML2
SAM9
SRI Braidwood
DRPIII
DRSIII
CAA1
LSL (electronic IRs only)
C. Pederson, DRS (hard copy - IRs only)
PLB1
TXN
ROPreports@nrc.gov (inspection reports, final SDP letters, any letter with an IR number)
U. S. NUCLEAR REGULATORY COMMISSION
REGION III
Docket Nos: 50-456; 50-457
Report No: 05000456/2007007 and 05000457/2007007
Licensee: Exelon Nuclear
Facility: Braidwood Station, Units 1 and 2
Location: Braceville, Illinois
Dates: March 12 through May 1, 2007
Inspectors: D. Smith, Project Engineer - Team Lead
N. Valos, Senior Operations Engineer
D. Jones, Reactor Engineer
M. Perry, Illinois Emergency Management Agency
Approved by: R. Skokowski, Chief
Branch 3
Division of Reactor Projects
Enclosure
SUMMARY OF FINDINGS
IR05000456/2007007, 05000457/2007007; 03/12/2007 - 05/01/2007; Braidwood Station,
Units 1 and 2. Identification and Resolution of Problems.
This report covers an approximate 16 week period of inspection by a project engineer, two
regional specialists, and an Illinois Emergency Management Agency inspector. Two Green
findings, which were both Non-cited violations, were identified by the inspectors. The
significance of most findings is indicated by their color (Green, White, Yellow, Red) using
Inspection Manual Chapter 0609, Significance Determination Process (SDP). Findings for
which the SDP does not apply may be Green or be assigned a severity level after NRC
management review. The NRCs program for overseeing the safe operation of commercial
nuclear power reactors is described in NUREG-1649, Reactor Oversight Process, Revision 3,
dated July 2000.
Identification and Resolution of Problems
In summary, the inspectors determined that the stations corrective action program was
effectively implemented as evidenced by the identification of plant issues through various
methods including departmental assessments and nuclear oversight audits. Plant issues
were documented in the stations corrective action program in a timely manner, and the
licensee generally implemented effective corrective actions to address plant issues and
events. The station has been effectively utilizing operating experience to prevent events
and improve performance at the station. However, an example of an inadequately performed
extent of condition review resulted in a Non-Cited violation during this inspection. A similar
problem with the licensees extent of condition reviews was also identified during the
October 2005 Problem Identification and Resolution Inspection.
The presence of a challenging nuclear oversight organization was apparent at the station.
This organization as well as other internal and external groups, have noted continuing
deficiencies in supervisory oversight. This issue with supervisory oversight was evident in the
licensees ability to sustain improved performance in several struggling areas, such as, the
control of transient combustibles, maintenance of personnel qualifications, and events related
to the bumping of plant equipment. The inspectors noted good communications of and
execution of the stations employee concern program. Additionally, the results from interviews,
conducted by the inspectors, reflected a safety conscious work environment at the station.
A. NRC-Identified and Self-Revealing Findings
Cornerstone: Mitigating Systems
- Green. The inspectors identified a violation of Technical Specification 5.4.1,
"Procedures," for the licensees failure to provide an adequate procedure to
ensure the continued operation of the "A" residual heat removal pump, during
cold leg recirculation mode of operation, during conditions when the "B" residual
heat removal pump was not available. The licensee initiated an issue report to
2 Enclosure
track the resolution of this finding. Subsequently, the licensee revised the
affected procedure on May 21, 2007 to ensure one residual heat removal pump
remained operable.
The licensees failure to maintain an adequate procedure to ensure the
continued operation of the A residual heat removal pump was more than minor
because the finding affected the mitigating systems cornerstone objective of
ensuring the availability and reliability of the emergency core cooling system to
respond to initiating events to prevent undesirable consequences. Specifically,
the finding was associated with the mitigating systems attribute of procedure
quality. The finding is of very low safety significance because the finding
screened as Green during the Phase 1 Significance Determination Process.
(Section 4OA2.a.1)
- Green. The inspectors identified a Non-Cited Violation of 10 CFR Part 50,
Appendix B, Criterion XI, "Test Control," because the licensee failed to include
several manual component cooling water system valves, that were required to
perform a safety function, in the inservice testing (IST) program and
subsequently test the valves in accordance with IST program requirements.
The finding was related to the cross-cutting area of Problem Identification and
Resolution. A cross-cutting aspect in the corrective action program was
identified because the licensee did not conduct an adequate extent of condition
review, for a previously missed IST surveillance on several essential service
water system valves. As a result, the licensee failed to identify that the
component cooling water systems valves required inclusion in and testing by
the IST program. The licensee initiated an issue report to track the corrective
actions for this finding. Subsequently, the licensee placed the valves on the
Plan-Of-The-Day Meeting Agenda to ensure testing, which was scheduled for
June 30, 2007.
The failure to account for these valves in the IST program was more than minor
because the finding affected the mitigating systems cornerstone objective of
ensuring the availability and reliability of the component cooling water and
residual heat removal systems when required to respond to initiating events to
prevent undesirable consequences. Specifically, the finding was associated with
the mitigating systems attribute of equipment performance. The finding is of
very low safety significance because the finding screened as Green during the
Phase 1 Significance Determination Process. (Section 4OA2.a.2)
B. Licensee-Identified Violations
None.
3 Enclosure
REPORT DETAILS
4OA2 Problem Identification and Resolution (PI&R) (71152B)
a. Assessment of the Corrective Action (CA) program
(1) Inspection Scope
The inspector reviewed the licensees CA program implementing procedures and
attended CA program meetings to assess the implementation of the CA program
by site personnel.
The inspectors reviewed risk and safety significant issues in the licensees CA program
since the last NRC PI&R inspection in October 2005. The selection of issues ensured
an adequate review of issues across NRC cornerstones. The inspectors used issues
identified through NRC generic communications, department self assessment, nuclear
oversight audits, operating experience reports, and NRC documented findings as
sources to select issues. Additionally, the inspectors reviewed issue reports generated
as a result of station personnels performance in daily plant activities. In addition, the
inspectors reviewed Issue Reports (IRs) and a selection of completed investigation from
the licensees various investigation methods, which included root cause, apparent
cause, equipment apparent cause, common cause, and quick human performance
investigations.
The inspectors selected four high risk systems, which included the emergency diesel
generator, circulating water, pressurizer, essential service water systems, to review in
detail. The inspectors review was to determine whether the licensee was properly
monitoring and evaluating the performance of these systems through effective
implementation of station monitoring programs. A five year review on the pressure
boundary, and essential service water systems was also undertaken to assess the
licensees efforts in monitoring for system degradation due to aging aspects. The
inspectors also performed partial system walkdowns of all the systems except for the
pressure boundary system due to the systems inaccessibility.
During the reviews, the inspectors determined whether the licensees actions were in
compliance with the stations corrective action program and 10 CFR 50, Appendix B
requirements. Specifically, the inspectors determined if station personnel was
identifying plant issues at the proper threshold, entering the plant issues into the
stations CA program in a timely manner, and assigning the appropriate prioritization for
resolution of the issues. The inspectors also determined whether the licensee assigned
the appropriate investigation method to ensure the proper determination of root,
apparent, and contributing causes. The inspectors also evaluated the timeliness and
effectiveness of corrective actions for selected IRs, completed investigations, and NRC
findings including non-cited violations.
This inspection constitutes one biennial sample of problem identification and resolution
as defined by Inspection Procedure 71152.
4 Enclosure
(2) Assessment
.1 Identification of Issues
The licensees effectiveness in implementing the stations CA program was
evidenced by the engagement of station personnel, from all departments, in
generating issue reports, documentation of findings by both internal and external
groups, number of self-identified trends, and results from the stations 2007 PI&R
Focused Area Self Assessment (FASA). Generally, department assessments and
nuclear oversight audits properly characterized issues as deficiencies when the
requirements of a CA program element were not met. Concurrently, documented
issues, meeting CA program element requirements, were appropriately specified as
recommendations to further improve station performance. However, the inspectors
did note one instance where a Nuclear Oversight (NOS) audit mischaracterized the
corrective action to revise a plant support procedure, to improve its quality, as a
recommendation. The audits characterization should have been specified as a
deficiency because the procedure could not be performed as written. This
inadequate procedure was not a violation of NRC requirements.
Based on the wide range of plant deficiencies and enhancements noted in IRs, the
inspectors determined that station personnel utilized the appropriate threshold level for
entering plant issues in the CA program. Additionally, maintenance rule, system health,
surveillance, and boric acid station program owners were appropriately generating issue
reports when program requirements were not met or upon the identification of adverse
trends. During, the inspectors reviewed control room logs from March 16 through
March 19, 2007, they noted, as did the sites PI&R FASA, that the control room
operators had not consistently generated IRs from documented operational issues or
equipment failures described in the logs. The licensee issued Operations Memo 1-07,
in January 2006, describing that such log entries warranted an issue report. In addition,
there were a few instances where the licensee did not generate timely issue reports
based on the inspectors observations. For example, a radiation area warning gate was
found blocking a painted warning sign on the floor warning personnel that radio use in
the area was prohibited. The licensees failure to ensure warning signs were not
blocked was considered minor as there were no adverse safety consequences as a
result of this failure. The gate was moved but the licensee did not generate an issue
report until several requests were made by the inspectors. Although the licensees had
not fully implemented the corrective actions from the PI&R FASA deficiency, the
inspectors considered the licensees progress slow to ensure IRs were consistently
generated.
The inspectors determined that the station implemented effective corrective actions to
address the causes of maintenance department personnels lack of involvement in
writing issue reports; this issue was identified during the 2005 NRC PI&R Inspection
(05000456/2005012; 05000457/2005012). The CA Program Manager conducted
presentations on the IR initiation process to ensure the maintenance staff understood
the process and computers were located in the maintenance shop for ease of access.
The results from the partial system walkdowns, conducted by the inspectors, indicated
that systems were well maintained and that identified deficiencies, such as oil and water
leaks, were entered into the CA program. During the walkdown, the inspectors did
5 Enclosure
identify a couple of deficiencies that had not been entered into the CA program,
however these deficiencies were minor in nature and did not adversely impact system
operability.
Findings and Observations
Inadequate Procedure to Ensure the Continued Operation of the "A" Residual Heat
Removal (RHR) Pump While in the Cold Leg Recirculation Mode of Operation
Introduction: The inspectors identified a finding of very low safety significance and
associated Non Cited Violation (NCV) of Technical Specification 5.4.1, "Procedures,"
for the licensees failure to provide an adequate procedure to ensure the continued
operation of the "A" RHR Pump, while in the cold leg recirculation mode of operation,
when the "B" RHR Pump was not available.
Description: On March 22, 2007, the inspectors identified that Braidwood Normal
Operating System Procedure, BwOP CC-8, "Isolation of CC Between Units 1 and 2,"
Revision 18, was inadequate. This procedure was used to support actions in the
Braidwood Emergency Operating Procedures (EOP), 1BwEP ES-1.3, "Transfer to
Cold Leg Recirculation Unit 1," Revision 104, and 2BwEP ES-1.3, "Transfer to Cold
Leg Recirculation Unit 2," Revision 104. The inspectors concluded that procedure
BwOP CC-8 was inadequate because the procedure would not ensure continued
operation of the safety injection and the centrifugal charging pumps when the
"B" RHR pump was not available.
Specifically, operators would use Procedure BwOP CC-8 for an event that required
the transfer of the emergency core cooling system to the recirculation mode due to
low-low refueling water storage tank level. Each unit specific EOP (1BwEP ES-1.3
and 2BwEP ES-1.3) specified the actions to complete the transfer to the recirculation
mode. Once in the recirculation mode, the centrifugal charging pumps and the safety
injection pump pumps were started using steps 1 through 6 of either EOP. Step 10.c,
of each EOP, specified aligning component cooling water system for post-loss of cooling
accident recovery using BwOP CC-14, "Post Loss of Cooling Accident [LOCA]
Alignment of the Component Cooling [CC] System, Revision 14. Procedure BwOP
CC-14 required the use of BwOP CC-8, "Isolation of Component Cooling [CC] Between
Units 1 and 2," for completing this task.
Procedure BwOP CC-8 was used to separate Unit 1 component cooling water system
flow from Unit 2 CCW flow during both normal and accident conditions. Therefore, if the
common (CC) heat exchanger was initially aligned to the unit experiencing a LOCA, the
operators were directed to secure the A RHR pumps. Steps to secure the pump were
specified by BwOP CC-8, Step F.1.c.6 for Unit 1 and Step F.2.c.7 for Unit 2, while the
RHR pumps were providing the water source supply to the safety injection and
centrifugal charging pumps (Piggyback Mode). The pump would be secured based on
the execution of either step, while manipulating several component cooling system
valves, during the time to align the common CC heat exchanger to the unit with the
LOCA. As a result of securing the only running residual heat removal pump
BwOP CC-8, while in the "piggy back" mode of operation, irreversible pump
damage could occur to the safety injection and both centrifugal charging pumps.
6 Enclosure
Upon the identification of this issue by the inspectors, the licensee initiated IR 00611024
to track this finding for resolution.
Analysis: The inspectors determined that the failure to provide an adequate procedure
to ensure the continued operation of the "A" residual heat removal pump, while in the
cold leg recirculation mode of operation, when the "B" RHR Pump was not available was
a performance deficiency warranting a significance evaluation. The inspectors reviewed
this issue against the guidance contained in Appendix B, "Issue Dispositioning
Screening," of Inspection Manual Chapter (IMC) 0612, "Power Reactor Inspection
Reports." The inspectors determined that the finding was more than minor in
accordance with IMC 0612, Appendix B, "Issue Disposition Screening," because the
finding affected the mitigating systems cornerstone objective of ensuring the availability
and reliability of the ECCS to respond to initiating events to prevent undesirable
consequences. Specifically, the finding was associated with the mitigating systems
attribute of procedure quality.
The inspectors evaluated the finding using IMC 0609, Significance Determination
Process,@ Appendix A, Significance Determination of Reactor Inspection Findings for
At-Power Situations,@ Attachment 1, dated March 23, 2007. The inspectors answered
No to all five questions under the Mitigating System Cornerstone column of
Attachment 1. The finding was not a design or qualification deficiency confirmed not to
result in loss of function per Generic Letter 91-18; did not represent a loss of system
safety function; did not represent an actual loss of safety function of a single train for
greater than its Technical Specification allowed outage time; did not represent an actual
loss of safety function of one or more non-Technical Specification trains of equipment
designated as risk-significant per 10 CFR 50.65 for greater than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />; did not
screen as potentially risk significant due to a seismic, flooding, or severe weather
initiating event.
Also, the finding did not affect the safety function of the high pressure recirculation
system unless one train of the residual heat removal system was initially failed or
unavailable due to maintenance. The unavailability of one safety system train was
inherently accounted for in the SDP. Therefore, when the operators would have
secured the residual heat removal pump per procedure, the loss of this train would have
been accounted for in the SDP. In addition, the regional Senior Reactor Analyst (SRA)
performed a confirmatory analysis to assess the risk of the finding using the site-specific
Braidwood Standardized Plant Analysis Risk Model, Revision 3.21. The SRA assumed
that if one train of residual heat removal system was unavailable that the high pressure
recirculation function would be failed because of the inadequate procedure. Using this
assumption, the SRA determined that the change in core damage frequency due to the
finding was less than 1.0E-6/yr, which was considered to be of very low safety
significance (Green).
Enforcement: Technical Specification 5.4.1, "Procedures," required, in part, that
written procedures be established, implemented, and maintained covering the
emergency operating procedures (EOPs) required to implement the requirements of
NUREG-0737, "Clarification of TMI Action Plan Requirements," and NUREG-0737,
Supplement 1. Item I.C.1 of NUREG-0737 and NUREG-0737, Supplement 1, Section 7,
required, in part, the development of EOPs to cover transients and accidents including
7 Enclosure
an event that required transfer of the emergency core cooling system to the cold leg
recirculation mode of operation.
Contrary to this requirement, on March 22, 2007, the inspectors discovered that
Braidwood Normal Operating Procedure, BwOP CC-8, "Isolation of CC Between Units 1
and 2," Revision 18, was inadequate. This procedure was used to support actions in
each unit specific Emergency Operating Procedure, 1BwEP ES-1.3, "Transfer to Cold
Leg Recirculation Unit 1" Revision 104, and 2BwEP ES-1.3, "Transfer to Cold Leg
Recirculation Unit 2," Revision 104. A procedural step failed to ensure the continued
operation of the safety injection and centrifugal charging pumps, while in the cold leg
recirculation mode of operation, when the "B" residual heat removal system pump was
not available. The licensee generated an issue report to followed up on the corrective
actions for this finding. Subsequently, the licensee revised BwOP CC-8 on
May 21, 2007, to ensure a RHR pump remained operable.
Because the finding is of very low safety significance and it was entered into the
licensees corrective action program (IR Number 00610994), the finding is being
treated as a NCV consistent with Section VI.A.1 of the NRC Enforcement Policy.
(05000455/2007007-01;05000456/2007007-01)
.2 Prioritization and Evaluation of Issues
The inspectors concluded that the licensee had properly prioritized issues based on
the safety significance of issues, and that issues were generally well evaluated. The
inspectors did not identify any issue reports that were not properly prioritized. In
addition, the inspectors observed several station ownership committee (SOC) and
management review board committee (MRC) meetings, and concluded that both
committees generally ensured the proper prioritization and appropriate investigation
assignments for plant issues. However, the inspectors did note several instances
where the oversight provided by both committees was not thorough. The sites 2007
PI&R FASA documented issue with the performance of the MRC. The inspectors
observed that the licensee had initiated the appropriate subsequent actions to evaluate
adverse trends. Due to the effective trending at the station, the inspectors did not
identify any adverse trends that had not been previously captured in the CA program
through department self-identification, NOS activities, quarterly nuclear safety review
board site visits, and the efforts of the sites PI&R FASA team.
The inspectors determined that the licensees selection of investigation methods, in
addressing site issues in all areas of plant operations, was appropriate and
commensurate with the safety significance of the issue or event. Also, the inspectors
determined that during extent of condition reviews, for plant issues, the licensees
reviews were generally adequate. However, this inspection as well as October 2005
NRC PI&R inspection identified shortcomings in the licensees extent of condition
reviews. The 2005 PI&R documented that a root cause evaluation, which was
associated with the precipitation of calcium carbonate in the ultimate heat sink
(IR199206), was too narrowly focused and failed to identified all the potentially affected
equipment. During this inspection, the licensee was again too narrow in umbrelling the
8 Enclosure
potentially affected components. In this case, the licensees extent of condition review
failed to identify that several component cooling water valves had not been included in
and tested in accordance with the requirement of the inservice testing (IST) program.
Regarding the licensees review of equipment operability, the inspectors determined,
that in general, they were appropriate with some shortcomings noted. The sites PI&R
FASA as well as the inspectors identified that some operability evaluations did not
discuss the affect on system operability for component failures identified when the
system was not required to be operable. An example noted by the inspectors, where
the operability basis was lacking, involved issue report Number 552355, "Relief valve
removed from 2SI8848 failed final seat leakage." The evaluation only documented that
an active leak was noted last cycle when the relief valve was installed. However, the
evaluation did not discuss how the system would have been affected if the valve had
lifted and subsequently experienced excessive seat leakage. Additionally, the residents
had been identifying similar examples during inspections and discussed these issues at
prior exit meetings with the licensee. The licensee generated IRs as a result of their
own FASA, the NRC PI&R inspection, and concerns from the resident inspectors in this
area.
Both the sites 2007 PI&R FASA and the inspectors, noted a few isolated instances,
where the licensees performance did not meet the requirements of several of these
CA program elements. The issues identified by the inspectors are documented below.
Findings and Observations
Component Cooling Water (CCW) System Valves not Included in the IST Program
Introduction: The inspectors identified a Green finding involving a Non-Cited Violation of
10 CFR Part 50, Appendix B, Criterion XI, Test Control, for the licensees failure to test
several manual component cooling water system valves, that required manipulation to
support the stations safety analysis, as specified by the inservice testing program.
Description: On March 18, 2007, the inspectors assessed the licensees extent
of condition review that had been performed for Inspection Report (IR)
Number 00522178. The IR was associated with the licensees failure to include
certain essential service water system valves in the IST program; the licensee
completed an apparent cause evaluation for this issue on August 21, 2006. In
assessing the quality of the extent of condition review, the inspectors identified that
certain manual CCW valves, which required manipulation during the transfer of the
emergency core cooling system to the recirculation mode, were not included in and
tested by the IST program.
Section 4.4.3, Manual Valves, of NUREG-1482, Guidelines for Inservice Testing
at Nuclear Power Plants, Revision 1, January 2005, specified that manual valves
credited in the licensees safety analysis to perform a specific safety function in
shutting down the reactor to a safe shutdown condition, maintaining the safe
shutdown condition, or mitigating the consequences of an accident be included in
the IST program. The inspectors identified that Braidwood Emergency Operating
Procedures, for both units, 1BwEP ES-1.3, Transfer to Cold Leg Recirculation Unit 1,"
9 Enclosure
Revision 104, and 2BwEP ES-1.3, Transfer to Cold Leg Recirculation Unit 2,"
Revision 104, specified eight manual CCW system valves in each procedure.
Operators were required to manipulate these valves, to meet the safety analysis
CC W water flow of 5000 gallons per minute (gpm) for the residual heat removal
heat exchangers, after an accident. Therefore, these CCW valves required testing
in accordance with the IST program; however, the licensee did not test the valves
because the valves were not included in the program. The licensees corrective
action for this issue included generating issue report Number 00610994 and placing
the valves, at least 16 valves between both units, on the Plan-Of-The-Day Meeting
Agenda to ensure the testing which, was scheduled for June 30, 2007.
Analysis: The inspectors concluded that the licensees failure to include the
component cooling water system valves in the IST program and subsequently
test the valves per IST program requirements was a performance deficiency
warranting a significance evaluation. The inspectors reviewed this finding
against the guidance contained in Inspection Manual Chapter (IMC) 0612,
Power Reactor Inspection Reports, Appendix B, Issue Dispositioning
Screening, dated November 2, 2006. The inspectors determined that the
licensees failure to test the component cooling water system valves in
accordance with the IST program was more than minor because the finding
affected the mitigating systems cornerstone objective of ensuring the availability
and reliability of the component cooling and residual heat removal systems.
Specifically, the finding was associated with the mitigating systems attribute of
equipment performance.
The inspectors evaluated the finding using Inspection Manual Chapter 0609,
Significance Determination Process,@ Appendix A, Significance Determination
of Reactor Inspection Findings for At-Power Situations,@ Attachment 1, dated
March 23, 2007. The inspectors answered No to all five questions under the
Mitigating System Cornerstone column of Attachment 1. The finding was not a
design or qualification deficiency confirmed not to result in loss of function per
Generic Letter 91-18; did not represent a loss of system safety function; did not
represent an actual loss of safety function of a single train for greater than its
Technical Specification allowed outage time; did not represent an actual loss of
safety function of one or more non-Technical Specification trains of equipment
designated as risk-significant per 10 CFR 50.65 for greater than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />; did not
screen as potentially risk significant due to a seismic, flooding, or severe weather
initiating event. Therefore, the issue screened as having very low safety significance
(Green). The finding was related to the cross-cutting area of Problem Identification
and Resolution and had the cross-cutting aspect of corrective action because during
the extent of condition review for IR Number 00522178, the licensee failed to identify
that at least 16 CCW system valves required inclusion in and testing by the
IST program.
Enforcement: Part 50 of 10 CFR, Appendix B, Criterion XI, "Test Control," states,
in part, that a test program shall be established to assure that all testing required to
demonstrate that safety-related structures, systems, and components will perform
satisfactorily in service is identified and performed in accordance with written test
procedures.
10 Enclosure
Contrary to this, on August 21, 2006, the licensees test program failed to ensure testing
of at least 16 safety-related component cooling water system valves, to demonstrate
that the valves would perform satisfactorily in service. This was due to an inadvertent
omission of these valves in the inservice test program. The licensees corrective action
for this issue included generating an issue report and placing the valves on the plan-of-
the-day meeting agenda to ensure testing of the valves.
Because the finding is of the very low safety significance and it was entered into the
licensees corrective action program (IR 00610994), the finding is being treated as a
non-cited violation (NCV) consistent with Section VI.A.1 of the NRC Enforcement Policy.
(NCV 05000455/2007007-02; 05000456/2007007-02)
LACK OF THOROUGHNESS BY THE MANAGEMENT REVIEW COMMITTEE
(1) IR Number 00496552 and IR Number 00507945, 1B First Stage Reheater
Drain Tank Hi-2 Alarm Failed to Reset During Reheater Valve/Intercept
Valve Surveillance
On June 1, 2005, IR Number 00496552 was written to identify a reactor power
excursion to 100.4 percent during the performance of Braidwood Operations
Surveillance Procedure, 1BwOS Technical Requirement Manual [TRM] 3.3.g.3,
Unit 1 Turbine Overspeed Protection Systems Valve Stem Freedom Check
(RV-IV Cycling). The power excursion occurred because the 1B First Stage
Reheater Drain Tank Hi-2 level alarm failed to reset, during the performance of
the monthly surveillance. When this failure occurred, increased steam flow was
routed to the condenser as a result of the emergency level control valve opening
to the condenser.
On July 10, 2005, IR Number 00507945 documented the occurrence of another
reactor power excursion event. The power excursion was due to the same
cause, failure of the 1B First Stage Reheater Drain Tank Hi-2 level alarm to reset
during the performance of the surveillance, as the June 2005 event. During the
second power excursion event, reactor power rose to 100.36 percent.
In response to these two events, the licensee closed the first event
(IR Number 00496552) to work order 900832 to repair and calibrate the
Hi-2 level switch during the next refueling outage (A1R13). Also, the licensee
performed a Quick Human Performance Investigation for the second event
(IR Number 00507945) and identified that both the station ownership committee
and the management review committee failed to realize that the first event was a
reactivity management event. The inspectors review also determined that the
Quick Human Performance Investigation failed to identify that the station
ownership committee and the management review committee did not identify the
need for timely interim correction action following the first event. Therefore,
appropriate actions were not taken, but could have been taken to prevent the
July 2005 power excursion event. The failure to take corrective actions to
prevent recurrence in this case was not required by NRC regulations, so no
violation occurred.
11 Enclosure
(2) Delayed Removal of Transient Combustibles
The inspectors attended the management review committee meeting on
March 13, 2007. Issue report Number 601183, Transient combustibles stored
at lake screen house without required permit, was reviewed by the committee
members. The issue report documented that the fire marshall was contacted for
this issue on March 6, 2007, and indicated that the quantify of transient
combustibles may be considered minor and that a permit may be required per
OP-AA-201-009, Control of Transient Combustible Materials, Revision 5.
The IR further documented that the transient combustibles would be removed
the following date. The inspectors were particularly concerned with the
disposition of this issue report due to recent and repeat transient combustibles
problems experienced at the station (see Section 4OA2.a.3 of this report). The
inspectors were concerned because the issue report did not document whether
the quantity of material was actually minor, a permit was required and had been
initiated to allow storage of the combustible, or other interim compensatory
measures had been established to allow the transient combustibles to remain at
the location.
During followup discussions with the licensee, the inspectors were informed that
significant followup discussions had taken place at the plan-of-the-day meeting
for this issue. The discussions indicated that the quantity of material was
exempt from procedural requirements; therefore, the licensee was in compliance
with the requirements of OP-AA-201-009. The inspectors were satisfied that the
licensee had appropriately recognized and assessed this issue, but, the
inspectors were concerned that the followup information was not subsequently
captured in the issue report. The omission of this information resulted in the
inability, of an independent review, to reach the same conclusion as the MRC
because of the missing information.
(3) Supervisor not held Accountability for Inadequate Radiological Briefing
The inspectors reviewed issue report Number 499359 which involved a worker
receiving an unexpected dose rate alarm. The inspectors review of the issue
report revealed that the duty supervisor was aware that the workers alarm may
alarm while performing a resin sluice evolution. However, the issue report did
not document that the duty supervisor involved with the briefing was held
accountable for conducting an inadequate briefing. The licensees failure to
address the specific performance of the supervisor was considered minor
violation since all radiation protection personnel was briefed on the issue. The
licensees corrective actions for this event was to brief all radiation protection
staff on this issue and the potential of a dose alarm for sluice activities.
(4) Cover Contamination Should have been Identified
The inspectors reviewed issue report Number 497545 which was associated with
a contamination event. The inspectors questioned the radiation protection
manager on why the contamination which had been discovered on the underside
12 Enclosure
cover plug on June 6, 2006, was not previously identified as a result of radiation
surveys. The radiation protection manager informed the inspectors that this
issue was a human performance issue. The licensees failure to have previously
identified the contamination was considered minor as the spread of
contamination was limited.
New Issue Report not Written for Expanded Scope of Deficient Supervisory
Performance In the Closure of Corrective Action Assignments
In IR Number 584820, the licensee documented that supervisory performance
with respect to corrective actions to prevent recurrence (CAPR) closure did not
meet procedural requirements. The licensee subsequently identified, through an
extent of condition review (EOC), that supervisory performance was not only
deficient for CAPR closure but also for other CA assignments closures that were,
associated with EACEs, CCA. The supervisors had also failed to provide
electronic approval for these types of CA assignment prior to the closure.
Although this new expanded deficiency of supervisory performance was part of
the EOC review, this information was captured in IR Number 584820. But, this
IR was limited to the inadequate closure of CAPRs assignments only.
The inspectors determined that the new information identified by the licensee
during the EOC review indicated that the deficient aspect of supervisory
oversight was of a larger magnitude than previously identified and documented
in the 2007 PI&R FASA. Therefore, it would have been prudent for the licensee
to have generated a new issue report at the time it was identified. However, the
licensee later recognized this issue as a missed opportunity during the FASA
and then the issue was captured in the CA program. Corrective actions taken by
the licensee included rerouting the CAPR to the department managers for
electronic approval. With respect to the other CA assignment types, the licensee
determined that department managers had approved all of the investigation type
assignments, therefore, the licensee determined not to obtain the electronic
approval for these assignments. Additionally, the licensee implemented interim
corrective actions to address these issues by inserting electronic route lists for all
open investigations, CAPR assignments, and EFR assignments through
March 31, 2007.
.3 Effectiveness of Corrective Action
The inspectors concluded that the licensee generally implemented corrective
actions that were effective in addressing plant issues. The licensees 2007 PI&R
FASA documented, isolated cases, where the requirements of this CA element
were not met. The licensee initiated issue reports for the specific PI&R FASA
deficiencies. The inspectors also noted minor examples where deficiencies were
noted in this CA element.
In general, the inspectors determined that the licensee had been identifying and
implementing corrective actions to arrest deficient plant performance.
Specifically, in January 2007, the licensee conducted a common cause analysis
(CCA), IR540986, as a result of an adverse trend in human performance errors
13 Enclosure
which had resulted in a number of station clock resets. The CCA identified two
common causes: 1) individuals were failing to detect and prevent human errors
through the use of basic human performance and technical human performance
error reduction tools; and 2) Management, including field supervision, was failing
to properly engage workers in the use of human performance and technical
human performance error reduction tools. The licensee developed a number of
corrective actions, some of which had been completed, which appeared
appropriate to address the common causes identified in the CCA.
Additionally, NOS and the nuclear safety review board noted that supervisory
oversight had been less than adequate in many instances. The licensee has
recognized that challenges, in a number of areas such as out-of-service errors,
qualification of site personnel, contractor injuries, control of transient
combustibles, and repetitive consequential bumping events. Specifically,
corrective actions associated with inadequate control of combustible materials
resulted in several repeat events. Ineffective corrective actions for a root cause
report in 2003 led to the NRC identification, in June 2006, of an NCV for
the failure to implement the licensee's procedure for control of combustible
materials. In February of 2007, the licensee again found that the root cause
evaluation initiated in 2006 was not timely and that interim actions were
ineffective at sustaining performance. Several mispositioning events have
occurred, during the last five years, due to plant personnel inadvertently bumping
plant equipment. These bumping incidents have affected safety or Technical
Specification systems and directly or indirectly caused unanticipated power
changes. The licensee has not characterized any of the events as a significant
condition adverse to quality; therefore, the licensee did not have to prevent the
occurrence of the events. In each event, the corrective actions were either
narrowly focused only on the particular system or limited to easily accessible
valves. An example of the licensees narrow corrective action involved the
mispostioning of a valve, on the day tank, for the 1B emergency diesel generator
system. The licensee had previously decided against altering this valve, to
address an earlier bumping event, on the basis that the valves location would
prevent inadvertent bumping on the valve. However, the licensee subsequently
removed the valves handle as corrective actions to this March event. Although
the licensee has implemented corrective actions to improve performance in
these struggling areas, the inspectors were concerned with the stations ability to
sustain performance in these areas.
Findings and Observations
Inadequate Response to the PI&R FASA Finding
The sites 2007 PI&R FASA documented that a corrective action assignment
from a root cause report was closed to a management assignment request
MREQ. These types of management assignments were not used as a corrective
action assignment type. The MREQ assignment action was to perform a
walkdown, in the plant, to identify equipment that may be susceptible to bumping
or other inadvertent manipulations. The inspectors reviewed the licensees
response to the documented FASA deficiency and concluded that the response
14 Enclosure
was inadequate because the licensees response specified that the use of an
MREQ assignment was acceptable because another CA assignment (IR526093-
59) was tracking the same MREQ assignment actions. Although, the licensees
response was not consistent with CA program requirements, the failure to use
the proper tracking assignment was considered minor since the issue was
captured in the CA program.
Inadequate Corrective Actions
The sites 2007 PI&R FASA documented that individuals (direct report to
managers) and managers did not properly close corrective action to prevent
recurrence (CAPRs) assignments. A supervisory review, was required by
procedure, prior to the supervisors direct reports closing the CAPRs. Although
the direct reports were to send their supervisors or managers the electronic
CAPR for approval, the supervisors failed to ensure the receipt of the electronic
CAPR. The inspectors concluded that the supervisors failure to implement this
procedural requirement was not addressed by the licensee. Specifically, the
licensees corrective action included documenting a Fundamental Management
Systems entry for the individuals, but the supervisors or managers did not
receive any Fundamental Management System entries. The inspectors
considered this corrective action narrowly focused and partially ineffective.
The inspectors determined that these corrective actions did not ensure that
supervisors and managers were held to the same level of accountability as
their direct reports even though both parties were responsible for the proper
closure of the CAPRs. Subsequently, the licensee made Fundamental
Management System entries for the supervisors.
Inappropriate Corrective Action Assignment
The inspectors reviewed two issues where the licensee assigned corrective
actions that did not appear appropriate. Specifically, cases where issues should
have been assigned as CA assignments, but were instead assigned as action
tracking item (ACIT). The ACIT assignments, as specified by procedural
guidance, track the completion of general actions required to address non-quality
related issues. Two examples of CAs for quality-related issues being tracked as
ACITs were noted. One issue was associated with the potential to adversely
impact safety related equipment, and the second issue could have resulted in not
meeting radiological posting requirements or labeling requirements, which are
necessary to inform workers of radiological hazards. The licensees failure to
use the proper CA program tracking assignment for these two issues were
considered minor because there were no adverse safety impacts as a result of
the use of the incorrect CA program assignment.
In the first case, the inspectors reviewed an item associated with a station worker
inappropriately modifying the plant that had the potential to render safety related
equipment inoperable. The licensee immediately removed the unauthorized
modification and determined that the modification did not adversely affect the
equipment while the unauthorized alteration was in place. The licensee initiated
appropriate corrective actions that entailed revising material utilized by the
15 Enclosure
worker and other in the same department. This material would be referenced in
the future when considering modifications to the plant. Also, the issue was
tailgated to the affected departments.
In the second case, the inspectors reviewed IR Number 499656 which was
associated with a NOS finding. The IR documented that NOS identified several
deficient radiological ropes, posting, and radioactive material labels. The RP
Manager generated an IR to understand why the issue had not been classified
as a CA assignment. The IR further documented that preventive maintenance
task, which was to have identified these types of issues, was inconsistently
implemented by radiation protection personnel. The corrective action specified,
was to ensure the preventive maintenance task addressed these issues and
address problems with label and rope degradation.
b. Assessment of the Use of Operating Experience (OE)
(1) Inspection Scope
The inspectors reviewed the licensees implementation of the station operating
experience program. Specifically, the inspectors reviewed implementing operating
experience program procedures, attended CA program meetings to observe the use
of OE information, completed evaluations of OE issues and events, and selected
2006 and 2007 monthly assessments of the OE composite performance indicators.
The inspectors review would determine whether the licensee was effectively integrating
OE experience in the performance of daily activities, evaluations of issues were proper
and conducted by qualified personnel, prevention of industry events, and use of
departmental assessments and NOS audits. The inspectors also assessed if corrective
actions, as a result of OE experience, were identified and effectively and timely
implemented.
(2) Assessment
The inspectors did not identify any findings of significance in this area. The inspectors'
review of operating experience reports identified that the licensee was appropriately
including the issues into the CA program and effectively implementing operating
experience at the station. During licensee staff interviews, the inspectors identified
that the use operating experience was considered during daily activities.
c. Assessment of Self-Assessments and Audits
(1) Inspection Scope
The inspectors assessed the stations ability to identify and enter issues into the station
CA program, prioritize and evaluate issues, and implement effective corrective actions,
through efforts from departmental assessments and NOS audits.
16 Enclosure
(2) Assessment
The inspectors concluded that the licensees departmental assessments and nuclear
oversight audits were effective at identifying plant deficiencies and enhancement
opportunities at an appropriate threshold level. Assessments and audits were thorough
and probing. The auditing and assessing teams were comprised on personnel with
appropriate skills, abilities, knowledge, and expertise, which resulted in the identification
of plant deficiencies, plant improvement recommendations, and plant strengths.
Assessments and audits properly characterized issues, and identified issues were
subsequently placed into the CA. One exception was noted as discussed under the
identification of issues section of this report. Also, the inspectors concluded that 2007
PI&R FASA was a very good effort that resulted in a quality product.
The 2007 PI&R FASA properly assessed each CA program element of the CA program
and determined that the stations CA program was effective. However, the PI&R FASA
documented that the requirements for various CA program elements, for a number of
isolated cases, were not met. The sites PI&R FASA noted that, although trending was
effective at the station, further improvement was needed with the performance of the
department CA program coordinators. As a result, the Site CA program Manager
increased the meeting frequency, with the department CA program coordinators from
once every two to three months to every two weeks. During these meetings, the Site
CA program Manager reviewed selected CA closure reviews for lessons learned and
departmental CA program trend boards/data for site improvements. The site CA
program Manager indicated that these efforts had resulted in improving the performance
of the department CA program coordinators during this short time period. Additionally,
the Site CA program Manager and the department CA program coordinators planned to
develop a CA program coordinators improvement action plan.
The inspectors assessment of the stations CA program was consistent with the results
documented in the site PI&R FASA. Primarily, the team identified additional examples
of the types of deficiencies, in the various CA program elements, that had been
identified during the stations PI&R FASA. In addition to similar minor issues identified
by the FASA and the NRC inspection teams, two more significant issues were identified
by the inspectors as evident by two NRC-identified violations and other observations
documented in the inspection report.
The inspectors held discussions with the station NOS manager to obtain a better
understanding of NOS activities with respect the underlying attribute of supervisory
oversight as a challenge to several problem areas and repeat issues. The NOS
manager indicated that the number one concern at the site was operations leadership in
driving performance at the station. As a result, configuration control and safety were
elements that the Operations Department was not setting the appropriate standards for
personnel performance; therefore, problems were occurring as a result of deficient
personnel performance. In addressing these issues, the licensee held a Site Wide
Standdown in February 2006, and performed a common cause analysis, for human
performance errors, which was reviewed by the inspectors.
17 Enclosure
d. Assessment of Safety Conscious Work Environment
(1) Inspection Scope
The inspectors assessed the stations safety conscious work environment through
the reviews of the stations employee concern program implementing procedures,
discussions with coordinators of the employee concern program, interviews with
personnel from various station departments, and reviews of issue reports. The
inspectors also reviewed the results from a Safety Culture Survey and the Braidwood
Occupation Safety and Heath Administration (OSHA) Voluntary Protection Program
survey.
(2) Assessment
The NRC 2005 PI&R had documented that a number of plant workers did not
understand the purpose of the stations Employee Concern Program (ECP).
Some workers were unaware that safety concerns could be raised through the ECP,
and other workers indicated that personal problems were addressed through the ECP.
The inspectors determined that the licensee appeared to have addressed this issue as
reflected by a very integrated ECP into station activities. The ECP coordinators were
very active in ensuring station awareness and understanding of the ECP. The
coordinators discussed the ECP at maintenance personnel alignment meetings and to
new station workers as they were hired at the plant. The results from the interviews of
plant staff conducted by the inspectors, it was evident that station personnel understood
the purpose of the ECP. Based on discussions with the site and backup ECP
Coordinators, the inspectors did not have any concerns with the implementation of the
ECP. The ECP Coordinators were properly implementing the site program by ensuring
workers identify were not revealed and properly monitoring the sites corrective action
program for issues which would be considered ECP concerns. Additionally, the
coordinator properly followed up on issues to ensure a chilled environment did not exist
at the station.
The results of the interview with station personnel indicated that plant workers were
knowledgeable about the tools available to them for raising nuclear safety concerns.
The inspectors did not receive any comments that indicated workers would be hesitant
about raising concerns. Furthermore, the workers communicated that station
supervision greatly supported the workers efforts in raising issues. None of the workers
indicated that they themselves or their co-workers had been retaliated against for raising
safety concerns.
The inspectors reviewed the licensees Midcycle Safety Culture Questionnaire
completed in December 2006. The results of the questionnaire indicated that a safety
conscious work environment was in existence at the station. The survey results were
consistent with comments made in the OSHA Voluntary Protection Program Survey
Results and by the Nuclear Safety Review Board (NSRB). The aspect of supervisory
oversight in the field was noted in the survey as an area needing improvement, and this
was consistent with the stations focused attention on supervisory oversight.
18 Enclosure
The inspectors also reviewed the OSHA Voluntary Protection Program Survey Results.
This survey was administered in August/September 2006 and sampled 100 percent of
station personnel. The inspectors review of the results indicated that station workers
primary concerns regarding personnel safety were that it appeared that personnel safety
took a backseat to outages and high workload times, supervisors did not spend enough
time in the plant, and personnel safety issues were not always resolved in a timely
manner. The licensee analyzed the comments and subsequently developed a list of
focus areas to address the concerns noted by the survey results.
4OA6 Meetings
.1 Exit Meeting
The inspectors presented the inspection results to Mr. D. Ambler and other members
of licensee management at the conclusion of the inspection on May 1, 2007. The
inspectors asked the licensee whether any materials examined during the inspection
should be considered proprietary. No proprietary information was identified.
An interim exit meeting was conducted on March 30, 2007, to discuss the
preliminary findings of the inspection with Mr. M. Smith and other members
of licensee management. No proprietary information was identified.
4OA7 Licensee-Identified Violations
No findings of significance were identified.
ATTACHMENT: SUPPLEMENTAL INFORMATION
19 Enclosure
SUPPLEMENTAL INFORMATION
KEY POINTS OF CONTACT
Licensee
K. Aleshire, Emergency Preparedness Manager
D. Ambler, Regulatory Assurance Manager
D. Burton, Licensed Operator Requalification Training Lead Instructor
M. Cichon, Licensing Engineer
L. Coyle, Maintenance Director
C. Dunn, Site Training Director
G. Golwitler, Site Corrective Action Program Manager
R. Leasure, Radiation Protection Technical Manager
J. Moser, Radiation Protection Manager
J. Petty, Regulatory Assurance
B. Schipiour, Work Control Director
M. Smith, Engineering Director
P. Summers, Nuclear Oversight Manager
T. Tierney, Chemistry, Environmental, and Radioactive Waste Manager
C. Walrath, Operations Shift Operations Supervisor
R. Wolen, Design Engineering Manager
Nuclear Regulatory Commission
R. Skokowski, Chief, Reactor Projects Branch 3
L. Kozak, Senior Reactor Analyst
Illinois Emergency Management Agency
C. Cecil, Head Resident Inspection, Nuclear Facility Safety Illinois Emergency Management
Agency
1 Attachment
LIST OF ITEMS OPENED, CLOSED AND DISCUSSED
Opened
05000451/2007007-01; NCV The licensees failure to maintain a procedure adequate
05000452/2007007-01 could have resulted in securing the only remaining residual
heat removal pump (Section 4.OA.2.a.1)
Closed
05000451/2007007-02; NCV Inadequate extent of condition review which failed to
05000452/2007007-02 identify that IST testing was not performed for component
cooling water systems valves (Section 4.OA.2.a.2)
Discussed
None
2 Attachment
LIST OF DOCUMENTS REVIEWED
ISSUE REPORTS GENERATED DUE TO THE INSPECTION
Walkdown
00606529; Minor Oil Leak on 1B SX Pump (1SX01PB); March 20, 2007
00606470; Small Shaft Packing Leak on 2B SX Strainer (2SX01FB); March 20, 2007
00609112; NRC PI&R IDD - Concern with MRC/SOC Response to IR 601183; March 26, 2007
00610562; PI&R IDD - RP Did not Write IR for IEMA Issues; March 29, 2007
00610161; NRC PI&R IDD Concerns with IR Operability Bases Statements; March 28, 2007
00610159; NRC PI&R IDD - CAS for M&TE Trend not Effective; March 28, 2007
00610259; PI&R IDD - Operating Log Entries not Referencing IR; March 28, 2007
00610497; NRC PI&R IDD - No CA to Address Inadequate Brief By RP supervisor;
March 29, 2007
00610499; NRC PI&R IDD - ACIT that Should Have Been a CA Assign (RP); March 29, 2007
00610507; NRC PI&R IDD - RP Accountability not Addressed for Survey; March 29, 2007
00606470; Small Shaft Packing Leak on 2B SX Strainer (2SX01FB); March 20, 2007
00607571; PI&R IDD - Enhancement Opportunity for BwOP CC-8; March 22, 2007
00609263; NRC Questions Why CC Valves in EOP are Not in IST Program; March 27, 2007
00610437; IST Bases Document Requires Revision; March 29, 2007
00610514; NRC PI&R IDD - OE Not Used/Reference in Operations ACE; March 29, 2007
00610994; NRC Potential Green NCV - Manual CC Valves Not in IST Program;
March 30, 2007
00611024; NRC Potential TBD Finding - Procedure BwOP CC-8 Inadequate; March 30, 2007
00610051; PI&R IDD - Typographical Errors in Completed CCA; March 27, 2007
00610514; NRC PI&R IDD - OE Not Used/Referenced in Operations ACE; March 29, 2007
ISSUE REPORTS REVIEWED DURING INSPECTION
Operations
00186275; Repeat Maintenance - 2CC9486 Failed Second LLRT (First Rework);
November 13, 2003
00353841; Procedure Needed for Swapping U-O CC Heat Exchanger; July 16, 2005
00367473; Potential Enhancements to Strainer Backwash Response; August 27, 2005
00383463; Evaluation of 1RY8028 LLRT Results; October 7, 2005
00434456; Recommendations from 2005 Reactivity Management FASA;
December 15, 2005
00435424; FASA Identified Procedure Change OP-AA-300 / WC-AA-105; December 19, 2005
00441759; Operations Reactivity Management FASA 2006; January 13, 2006
00441919; Braidwood FASA - In-Service Testing (IST); January 13, 2006
00445423; Large Number of Consequential Configuration Control Events; January 24, 2006
00446241; OPEX Review - Lesson Learned from Byron IR 444685/444975; January 26, 2006
00446244; OPEX Review - Lesson Learned from Byron IR 444975/444499; January 26, 2006
3 Attachment
00452245; Unplanned LCO Entry on 2A DG Due to Low Temperature; February 10, 2006
00457480; Unit 2 RWST Level Slowly Decreasing; February 23, 2006
00479478; Unplanned LCO Entry into 3.4.12; April 17, 2006
00484683; 1RY8028 Body to Bonnet Steam Leak; April 29, 2006
00485871; A1R12 LL 16 percent Level Drop in 1A SI Accum During BwVSR 3.4.14.1;
May 2, 2006
00496552; 1B First Stage RDT HI-2 Alarm Failed to Reset During RV/IV SRV; June 5, 2006
00526168; Potential Unplanned LCO Due to Missed Technical Specification Surveillance;
August 31, 2006
00537802; Potential Trend in the In-Service Testing Program (IST); September 29, 2006
00507945; 1B First Stage RDT HI-2 Alarm Failed to Reset During RV/IV SRV; July 10, 2006
00523372; Unplanned LCO - Procedure Problem with 1B SSPS Bi-Monthly; August 24, 2006
00534719; Potential Missed Tech Spec Surveillance; September 22, 2006
00546312; 2CC9486 Fails Its A2R12 As Found Local Leak Rate Test; October 19, 2006
00547003; Valve Disc for 2CC9486 (A2R12 Outage); October 20, 2006
0554857; 2HD005A Acting Erratic; November 7, 2006
00555147; U-2 Reactor Power Effects During 2HD005A Restoration; November 18, 2006
00561698; NRC Concerns Identified During U-1 ECCS Vent and Valve Surveillance;
November 24, 2006
00568853; NRC IDD Concern with Past Operability of 2CC9486; December 13, 2006
00574749; Reactivity Management FASA - Standards Deficiency; January 3, 2007
00584642; PI&R FASA IDD - Concern with Response to SX Strainer Finding; January 29, 2007
00585093; PI&R FASA IDD - Concern with RCR Quality on 1CV243 Bump Event;
January 30, 2007
00585839; PI&R FASA IDD - Concern with FME Not Wholly Addressed in EACE;
January 31, 2007
00594356; Unit 1 Reactor Power Effects During 1MS040D Clearance Order; February 21, 2007
00468506; 1DO2006B Found Out of Normal Position; March 20, 2006
00330826; Valve 2RE004B Inadvertently Bumped; May 2, 2005
00472884; 1FW076 Failed Open Due to 1IA1234 Found 90 percent Closed; March 30, 2006
00095256; Unplanned Entry Into AAR 2BwOS PR-1a For Failure of 2RR08J; February 14, 2002
Operations log entries
Operator logs from 3/16/2007 through 3/18/2007
Operations department memorandum 1-07;Issue Generation and Log Keeping Clarifications;
February 20, 2007
00582356; PI&R FASA ID'd Operations Log Entries Without Log Entries; January 23, 2007
00607368; TLDs Were Put Thru X Ray Machine; March 22, 2007
Radiation Protection
IR485357; Contract Employee Exited PA after Alarming Rad Monitor; May 1, 2006
IR589515; Unacceptable RAD Shipping Practices; February 9, 2007
IR546010; Unexpected ED Alarm 377' IMB Near R24; October 19, 2006
IR528711; Shielding Blankets Need Adjustment to Shield Hot Spot; September 8, 2006
IR528700; RP Source Control Check-in Deficiency; September 5, 2006
IR517975; RAD Source found in Warehouse 15 (Good catch); August 8, 2006
IR515674; Extra RAD Exposure Taken due to Computer EPN Problem; August 1, 2006
IR486492; Relief Valve 1SD016 Isotopically Contaminated; May 4, 2006
4 Attachment
IR484195; A1R12 LL ALARA/Shielding Issues w/Westinghouse CRDM Equipment;
April, 28, 2006
499359; Unexpected Electronic Dosimeter Dose Rate Alarm; June 12, 2006
497545; Level 2 Personnel Contamination Event; June 6, 2006
499656; NOS ID: Several Outdoor Radiological Postings Found Unsat; June 13, 2006
547064; Unit 2 reactor Head New O-rings Damaged; October 20, 2006
589845; NOS ID: Negative Trend in the Lack of Rad-Waste Shipments; February 9, 2007
482383; Level 3 PCE (Westinghouse); April 22, 2006
518432; VCT Valve Aisle Contaminated During Fill and Vent; August 9, 2006
532256; Seavan Number 14 and 27 in degraded Condition; September 2007
528670; Pipe Requires Flush Flushing to Eliminate Hot Spot Number 59; September 8, 2006
528681; Pipe Requires Flush flushing to Eliminate Hot Spot Number 54; September 8, 2006
539072; Pipe Requires flush To Eliminate Rad Hot Spots; October 2, 2006
481427; A1R12 LL - RP Air Samples are not being Properly Prepared; April 20, 2006
536685; Elevated Dose Rates on U2 CV Letdown Piping; September 27, 2006
490604; Increasing Trend In Personnel Contamination Events PCE; May 16, 2006
523419; Possible Unmonitored Vent Path and Water Leakage U-2 CWA; August 24, 2006
243310; LHRA Left Unlocked and Unguarded-Tech Spec Violation; August 10, 2004
Maintenance
262233; 1SI8853B Failed to Repeat Required Lift Pressure; October 11, 2004
278269; Relief Valve Failed to Lift; December 1, 2004
329172; Valve Removed From 2SI8853B Failed Seat Leak & Lift Test; April 27, 2005
332762; IST Relief Valve 2SI8851 Failed Pressure Test; May 6, 2005
454898; Need WR For Contingency Work Order for 2SI8853A; February 16, 2006
459085; Spare Relief Valve From 1SI8853B Failed Testing; February 27, 2006
507260; Pre-Outage Task for 2SI8858 Performed Three Months Early; July 7, 2006
534034; Boron Identified on 1SI8851 During U01 ECCS Vent & Valve; September 24, 2006
546046; 2SI8851 Relief Valve Failed Initial Lift Test Low"; October 19, 2006
552355; Relief Valve Removed From 2SI8858 Failed Final Seat Leakage; November 2, 2006
553376; Relief Valve Removed From 2SI8842 Failed Initial Testing; November 4, 2006
554657; Safety Relief Valve Product Advisory"; November 7, 2006
565162; Relief Valve S/N N56877-00-0147 Failed Testing; December 4, 2006
576883; IST Relief Valve 1CS08MB Failed As Found Testing; January 9, 2007
102884; Pressurizer Safety Valves Test Out of Tolerance; April 8, 2002
Maintenance Adverse Trends
554034; Safety - Contractor Adverse OSHA Recordable Trend; November 6, 2006
476197; Safety Issue - Potential Negative Trend; April 7, 2006
385495; Trend - Unsafe work Practices with Working at Heights; October 13, 2006
584806; PI&R FASA IDD - Potential Adverse Trend With Safety; January 26, 2007
546473; A2R12 - Adverse Injury Trend; October 19, 2006
445525; Potential Adverse Trend in Safety; January 24, 2006
388537; Potential Trend: Safety Work Practices in Maintenance; October 21, 2005
540986; CCA, Potential Adverse Trend - Braidwood station Human Performance;
January 12, 2007
5 Attachment
Lost M&TE
448780; "M&TE Lost in 2005"; June 26, 2006
484738; "M&TE Being Declared Lost (EMD)"; April 29, 2006
484741; "M&TE Being Declared Lost (IMD)"; April 29, 2006
484742; "M&TE Being Declared Lost (MMD)"; April 29, 2006
484745; "M&TE Being Declared Lost (Venture)"; April 29, 2006
484757; "M&TE Being Declared Lost (Unknown work group)"; April 29, 2006
509129; "M&TE Being Declared Lost (MMD)"; July 13, 2006
509438; "M&TE Being Declared Lost (Unknown work group)"; July 14, 2006
509485; "M&TE Being Declared Lost (Operations)"; July 13, 20064
548931; "M&TE Being Declared Lost (EMD)"; October 25, 2006
548948; "M&TE Being Declared Lost (IMD)"; October 25, 2006
548979; "M&TE Being Declared Lost (MMD)"; October 25, 2006
548987; "M&TE Being Declared Lost (Venture)"; October 25, 2006
550954; "Potential Trend-Lost M&TE (Maintenance)", October 30, 2006
551583; "M&TE Being Declared Lost (Venture)"; October 31, 2006
551590; "M&TE Being Declared Lost (EMD)"; October 31, 2006
595587; "M&TE Being Declared Lost (Venture)"; February 24, 2007
595589; "M&TE Being Declared Lost (EMD)"; February 24, 2007
595591; "M&TE Being Declared Lost (Reactor Services)"; February 24, 2007
595592; "M&TE Being Declared Lost (Engineering Programs)"; February 24, 2007
595596; "M&TE Being Declared Lost (Engineering Work Group)"; February 24, 2007
Assessments and Audits/NOS/NSRB
595565; NOS ID PDR - Clearance and Tagging Program; February 24, 2007
394859; NOS PDR - Maintenance Unsafe Work Practices; November 4, 2005
496580; NOS IDD PDR Maintenance Performance; June 2, 2006
552654; NOS IDD Clearance and Tagging PDR; November 2, 2006
545330; NOS IDD Unchocked Cart in Turbine Building (venture); October 17, 2006
481167; NOS Identified an Adverse Trend on Control of Carts; April 20, 2006
196476-05; Source Accountability and Control Check-In Self Assessment; September 5, 2006
593726; NOS Elevation Notice Transient Combustible Material Control; February 28, 2007
574353; NOS ID FLS Qualification Documentation; February 2, 2007
560730; PI&R FAS; February 8, 2007
Braidwood Nuclear Safety Review Board Meeting; January 16 and 17, 2007
Braidwood Nuclear Safety Review Board Meeting; September 7 and 8, 2006
Braidwood Nuclear Safety Review Board Meeting; May 17 and 18, 2006
Braidwood Nuclear Safety Review Board Meeting; January 25 and 26, 2006
Braidwood Nuclear Safety Review Board Meeting; October 18 and 19, 2005
Braidwood Station Post Outage Review (A2R12); December 15, 2006
A1R12 Post Outage Review
Braidwood NOS Site Status Report; March 26, 2007
Braidwood NOS Site Status Report; December 18, 2006
Corrective Action Program
380114; EFR Determined CA 267878-08 to be Ineffective; September 30, 2005
463049; Missed Scheduled Training - Plant Engineering; March 7, 2006
432350; Missed ERO Training (Project Management); December 7, 2005
381339; 10/5/05 FLS Training Session Cancelled; October 2, 2005
6 Attachment
457631; Trend Identified in RP Department Clock Resets; February 23, 2006
588941; PI&R FASA IDD - Deficiency with Inadequate Closures of CAs; February 7, 2007
591311; NSO Reactivation Guide Contained Omissions; February 13, 2007
584820; Pi&R FASA IDD - CAPRs Closed Without Dept Mgr Approval; January 26, 2007
585093; PI&R FASA IDD - Concern w/RCR Quality on 1CV243 Bump Event; January 26, 2007
526093; NRC Potential NCV - Unplanned LCO entry Due to 1CV243 Bumped; August 31, 2006
584990; PI&R FASA IDD - Inadequate Closed CAPR for Procedure Rev.; January 26, 2007
604938; Outboard Bearing Housing Leaks on 2B SX Pump; March 13, 2007
502360; U-2 SAC has Hi Vibes; June 21. 2006
501537; U-2 SAC Tripped on Hi Oil Temp; June 19, 2006
397297; CAP Trend Code Application Deficiency; November 10, 2006
585860; PI&R FASA IDD - Trend with Failures on OLIS-CF084; January 26, 2007
585325; PI&R FASA IDD - Trend with U-2 SAC Emergent Problems; January 26, 2007
585778; PI&R FASA IDD - No CA Initiated for Common Cause (ops); January 26, 2007
394664; DC Bus 111 Ground Alarm Toggling; November 2005
586507; PI&R FASA IDD, Issues w/operability & Reportability Reviews; February 1, 2007
601183; Transient Combustibles Stored at LSH
560730; FASA Report for PI&R Pre-NRC Inspection; February 02, 2007
563100; NRC Identified NCV Issued-Control of Combustible Materials; November 29, 2006
399061; EP Drive-In Drill Failures; November 15, 2005
526093; NRC Potential NCV-Unplanned LCO Entry Due to 1CV243 Bumped; October 13, 2006
556692; Braidwood Unit 2 in Chemistry Action Level 2 for I131; November 11, 2006
426461; Increasing Trend in EFRS that are Collectively Ineffective; November 21, 2005
0116242; NRC Identified Issues in SX Pump Rooms; July 18, 2002
0154441; Potential 1RY456 Diaphragm Leak; April 13, 2003
0156624; Enhancements to PORV Accumulator Test - BwOSR 3.4.11.3; April 30, 2003
0324246; NRC Observations Noted on 2A SX Pump (2SX01PA); April 13, 2005
BRW MRC Agenda for Tuesday, March 13, 2007
BRW MRC Agenda for Wednesday, March 14, 2007
BRW SOC Agenda for Tuesday, March 13, 2007
Braidwood; A Day In The Plant Observation; January 10, 2007
Braidwood; A Day In The Plant Observation; November 14, 2006
Braidwood; A Day In The Plant Observation; August 1, 2006
Braidwood; A Day In The Plant Observation; June 27, 2006
Braidwood; A Day In The Plant Observation; February 2, 2006
586509; PI&R FASA IDD - Deficiency w/department CAPCO performance; February 1, 2007
Department CAPCO Indoctrination Guide
BwHS 4002-097; Surveillance of Air Duct Smoke Detectors 1/2XY-VT001; Revision 2
MA-BR-723-002; Smoke Detector Testing; Revision 2
Investigations
445423; Complete CCA for 2005 Configuration Control Events; March 31, 2006
148014; Common Cause Identified for FP Sensitive Issues; February 20, 2004
053104; Corrective Actions From CCA 77614; December 28, 2001
585093; PI&R FASA-Concern W/ RCR quality on 1CV243 Bump Event; January 30, 2007
583763; RCR-574353; TQ-AA-210-4303 Rev 1 not Implemented; January 26, 2007
591458; Root Cause Report Identified Poor Actions to NCV Finding; February 14, 2007
547064; Quick Human Performance Investigation Template; Unit 2 Reactor Head New O-Rings
Damaged
7 Attachment
Other
490604; Increasing Trend in Personnel Contamination events PCE; May 16, 2006
00184989; 2CC9486; (CC Supply to RCP Inside CIV) Failed LLRT; November 6, 2003
00265910; Inadvertent TRM LCOAR 3.4.C Entry Due to Excessive PZR C/D; October 21, 2004
00291106; Unplanned LCOAR Entry - 2B DG Jacket Water Circ Pump Trips; January 15, 2005
00306938; Unplanned LCO/Risk Change During ACB 1424 Trip Checks; March 1, 2005
00370649; 1FW039A Failed 1BwOSR 3.6.3.5.FW-3; September 6, 2005
03935515; Adverse Trend - Consequential Procedure Change Errors; November 2, 2005
00437222; Unplanned LCO Entry Due to OB VC Chiller Trip; December 28, 2005
00522178; Unplanned LCO Entry - Missed Surveillance Requirement; August 21, 2006
434566; Maintenance Audit, Maintenance Functional Area; March 17, 2006
308084; Bumping of Hi-2 Level Switch Causes Heater Isolation; March 3, 2005
321000; High and Hi-2 Level in 27A Heater; April 4, 2005
328487; Wrong Valve Body Installed for 2RH8427A in 2AR11; April 25, 2005
329241; Reactive Load Transients on 2B DG During PMT; April 27, 2005
454883; Shortfalls in Implementation of M&TE Program; March 30, 2006
474360; EMD Respirator Qualifications Below 50 percent; April 3, 2006
559573; Potential Adverse Trend-Braidwood Station Procedure Adherence; February 6, 2007
591444; Historical-Improperly Closed Action Items From 2003 RCR; February 14, 2007
00274721; HI-2 Isolation of 15-17 Heaters Causing OPDT Runback; November 18, 2004
00526093; NRC Potential NCV - Unplanned LCO Entry Due to 1CV243 Bumped;
August 31, 2006
00601545; Need WR to Disassemble and Inspect Valve 1CW018; March 9, 2007
DRAWINGS
M-66 Sheet 2; Diagram of Component Cooling Unit 1; Revision AO
M-66 Sheet 3A; Diagram of Component Cooling Units 1 and 2; Revision AU
M-66 Sheet 4D; Diagram of Component Cooling Units 1 and 2; Revision BC
M-139 Sheet 2; Diagram of Component Cooling Unit 2; Revision AI
REFERENCES
LS-AA-115; Operating Experience; Revision 10
LS-AA-120, Issue Identification and Screening Process, Revision 6
LS-AA-125, Corrective Action Program (CAP) Procedure, Revision 11
LS-AA-125-1001; Root Cause Analysis Manual; Revision 6
LS-AA-125-1002; Common Cause Analysis Manual; Revision 5
LS-AA-125-1003; Apparent Cause Evaluation Manual; Revision 7
LS-AA-125-1004; Effectiveness Review Manual; Revision 2
LS-AA-125-1005, Coding and Analysis Manual, Revision 5
LS-AA-126; Self-Assessment Program; Revision 5
LS-AA-126-1001; Focused Area Self-Assessments; Revision 4
LS-AA-126-1005, Check-In Self Assessments, Revision 3
LS-AA-126-1006, Benchmarking Program, Revision 1
EI-AA-1; Employee Issues; Revision 1
EI-AA-101; Employee Concerns Program; Revision 6
EI-AA-100-1003; Employee Issues Advisory Committee Notification; Revision 0
EI-AA-101-1002; Employee Concerns Program Trending Tool; Revision 3
MA-AA-716-017, Equipment Readiness and Reliability, Revision 1
Employee issues; E1-AA-1; Revision 1
8 Attachment
Employee Concerns Program; EI-AA-101; Revision 6
Employee Concerns Program Trending Tool; Revision 3; EI-AA-101-1002
EI-AA-101-1001; Employee Concerns Program Process; Revision 4
NOS Objective Evidence Report; May 9, 2006 - June 16, 2006
Nuclear Oversight Quarterly Report, NOSPA-BW-06-4Q; January 24, 2007
NO-AA-200-002-1002; Nuclear Oversight Audit Templates; Revision 6
NO-AA-200-002; Nuclear Oversight Regulatory Audit Procedure; Revision 10
TQ-AA-1018; Trainee Conduct Standards; Revision 1
BwAP 340-1; Use of Procedures for Operating Department; Revision 20
1BwEP ES-1.3; Transfer to Cold Leg Recirculation Unit 1; Revision 104
2BwEP ES-1.3; Transfer to Cold Leg Recirculation Unit 2; Revision 104
1BwGP 100-5; Plant Shutdown and Cooldown; Revision 33
BwOP CC-8; Isolation of CC Between Units 1 and 2; Revision 18
BwOP CC-14; Post LOCA Alignment of the CC System; Revision 11
OP-AA-106-101-1006; Operational and Technical Decision Making Process; Revision 4 Memo
No. BR-40, Expectations for Extending Issue Report Cause Investigations and Corrective
Action Due Dates, Revision 1
Procedure BwMP 3305-109; "IST and Non-IST Safety / Relief Valve testing; November 8, 2006
Procedure MA-AA-716-040; "Control of Portable Measurement and Test Equipment Program";
January 26, 2007
NRC Information Notice 2006-24; "Recent Operating Experience Associated With Pressure and
Main Steam Safety/Relief Valve Lift Setpoints; November 14, 2006
Memo No. BR-059, Day In The Plant Observation Program, Revision 1
Memo No. BR-055, Expectations for Root Cause Report Quality of Preparation, Oversight, and
Timeliness, Revision 0
NUREG-1482; Guidelines for Inservice Testing at Nuclear Power Plants; Revision 1
WCAP-12232; Commonwealth Edison Company - Byron/Braidwood Plants - Component
Cooling Water System; Revision 0
WCAP-13588, Operating Strategies for Mitigating Pressurizer Insurge and Outsurge
Transients; March 1993
Engineering
599516; OVA09FB - Carbon Sample Failed the Test; March 5, 2007
374437; Inaccurate ECC Calculation; September 16, 2005
465719; FRAC Tank Berm Collapse; March 13, 2006
480489; Boric Acid Accumulation at Bottom of PZR; April 19, 2006
484671; 1FW039 Repack Caused Decrease in 1D S/G Lvl; April 29, 2006
428868; Elevated Tritium Levels in On-Site Monitors; November 30, 2005
504769; 2CS01PA Seal Leakage During RTS; June 29, 2006
324966; Increased RCS Leakage Identified-Mispositioning of 2PR5045; April 15, 2005
523419; Possible Unmonitored Vent Path and water Leakage U-2 CWA; August 24, 2007
Security
IRs generated from June 2006 through March 9, 2007 in security
Miscellaneous
369873; Training Evaluation Methodology Deficiency; September 3, 2005
582168; CMO Group PQD Revision Issues"; January 23, 2007
583451; Chemistry ID'd Qualification Issue During Qual Reviews; January 25, 2007
9 Attachment
583465; Engineering Qualification Review Results; January 25, 2007
583480; Chemistry ID'd Missing Qualifications in PQD; January 25, 2007
583477; Chemistry Training Qualification Review; January 25, 2007
583484; Incomplete Training Documentation; January 25, 2007
585938; HR Identified Gaps In Quals In PQD; January 31, 2007
Operations Department Memorandum 06-6 "Inadvertent Contact Devices; September 26, 2006
10 Attachment
LIST OF ACRONYMS USED
ACIT Action Tracking Item
ADAMS Agency-Wide Document Access and Management System
CA Corrective Action
CAPR Corrective Action to Prevent Recurrence
CCA Common Cause Analysis
CC Component Cooling
CCW Component Cooling Water
CFR Code of Federal Regulation
EACE Equipment Apparent Cause Evaluation
ECP Employee Concern Program
EOP Emergency Operating Procedure
FASA Focused Area Self Assessment
DRS Division of Reactor Safety
EOC Extent of Condition
GPM Gallons Per Minute
IMC Inspection Manual Chapter
IR Issue Report
IST Inservice Testing
LOCA Loss of Coolant Accident
MRC Management Review Committee
NCV Non-Cited Violation
NOS Nuclear Oversight
NRC United States Nuclear Regulatory Commission
OE Operating Experience
OSHA Occupation Safety and Health Administration
PI&R Problem Identification and Resolution
QHPI Quick Human Performance Investigation
RCA Root Cause Analysis
SDP Significance Determination Process
SOC Station Ownership Committee
SRA Senior Reactor Analyst
TRM Technical Requirement Manual
11 Attachment