ML063550261
| ML063550261 | |
| Person / Time | |
|---|---|
| Site: | Indian Point  |
| Issue date: | 12/21/2006 |
| From: | NRC Region 1 |
| To: | |
| References | |
| EA-06-311 IR-06-006 | |
| Download: ML063550261 (28) | |
See also: IR 05000247/2006006
Text
U.S. NUCLEAR REGULATORY COMMISSION
REGION I
Docket No:
50-247
License No:
Report No:
Licensee:
Entergy Nuclear Northeast (Entergy)
Facility:
Indian Point Nuclear Generating Unit 2
Location:
295 Broadway, Suite 3
Buchanan, NY 10511-0308
Dates:
September 18 through October 6, 2006
Team Leader:
T. Walker, Senior Project Engineer, Division of Reactor Projects (DRP)
Inspectors:
M. Cox, Senior Resident Inspector, DRP
S. McCarver, Project Engineer, DRP
J. Benjamin, Resident Inspector, DRP
C. Long, Project Engineer, DRP
Observer:
S. Smith, Reactor Engineer, DRP
Approved by:
Eugene W. Cobey, Chief
Projects Branch 2
Division of Reactor Projects
Enclosure
ii
SUMMARY OF FINDINGS
IR 05000247/2006-006; 09/18/2006 - 10/06/2006; Indian Point Nuclear Generating Unit 2;
Problem Identification and Resolution.
This team inspection was performed by three regional inspectors and two resident inspectors.
Three findings of very low safety significance (Green) were identified, two of which were also
non-cited violations (NCVs). The significance of most findings is indicated by their color
(Green, White, Yellow, Red) using Inspection Manual Chapter (IMC) 0609, Significance
Determination Process (SDP). Findings for which the SDP does not apply may be Green or be
assigned a severity level after NRC management review. The 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
The inspectors concluded that the implementation of the corrective action program at Indian
Point Unit 2 was generally effective. The inspectors noted that Entergy staff had a low
threshold for identifying problems and entering them in the corrective action program. The
inspectors also noted that once entered into the system, items were screened, prioritized, and
evaluated commensurate with their significance using established criteria. The inspectors
determined that corrective actions addressed the identified causes and were typically
implemented in a timely manner. In addition, the team noted that Entergy was generally
effective in reviewing and applying lessons learned from industry operating experience. The
inspectors found that audits and assessments were critical and, in most cases, appropriate
actions were taken to address identified issues. However, the inspectors also found that the
results of an independent safety culture assessment were not entered into the corrective action
program for timely evaluation and appropriate action.
The inspectors found that most workers indicated that they would raise issues that they
recognized as nuclear safety issues. However, the inspectors also found that a number of
workers interviewed indicated that they were aware of individuals they perceived as having
been treated negatively by management for raising issues; most of these workers were in the
Instrumentation and Controls (I&C) department. Some workers expressed reluctance to raise
issues under certain circumstances due to a number of reasons, including fear of disciplinary
action and concerns with the efficacy of the corrective action program. While most workers
made a distinction between nuclear safety issues and other concerns, the inspectors noted that
some of the illustrative examples provided by plant workers could have nuclear safety
implications. However, the inspectors did not identify any more than minor issues, which had
not been raised.
There were two Green NCVs and one Green finding identified by the inspectors during this
inspection. One of the NCVs was associated with a failure to identify a condition adverse to
quality associated with the auxiliary feedwater (AFW) system. The second NCV was
associated with a failure to fully evaluate leakage into a steam generator. The finding was
Enclosure
iii
associated with the failure to enter adverse conditions into the corrective action program for
evaluation and appropriate action.
a. NRC Identified and Self-Revealing Findings
Cornerstone: Mitigating Systems
Green. The inspectors identified a non-cited violation (NCV) of 10 CFR 50, Appendix B,
Criterion XVI, Corrective Action, in that, Entergy failed to identify a condition adverse to
quality associated with improper internal clearances on BFD-68, an auxiliary feedwater
check valve, in the corrective action program. Specifically, upon inspection in
September 2006, the gasket between the valve's body to bonnet seal was found
over-crushed causing the gasket to partially unwind, potentially impacting valve
operation. Gasket damage was noted in work orders during internal valve inspections of
BFD-68 performed in 1997 and 2002; however, the deficiencies were not identified in
the corrective action program. Consequently, the problem was not evaluated and
corrected prior to reassembly of the valve. Entergy entered this issue into the corrective
action program, evaluated the condition, and conducted repairs to the valve to ensure
the proper gasket crush was obtained.
The inspectors determined that this finding was more than minor because it was
associated with the Equipment Performance attribute of the Mitigating Systems
cornerstone; and, it affected the cornerstone objective of ensuring the availability,
reliability and capability of systems that respond to initiating events to prevent
undesirable consequences. The inspectors evaluated the significance of this finding
using Phase 1 of IMC 0609, Appendix A, "Significance Determination of Reactor
Inspection Findings for At-Power Situations," and determined that the finding was of
very low safety significance because it was not a design or qualification deficiency; it did
not result in the loss of a system safety function or a train safety function for greater
than the Technical Specification Allowed Outage Time; and it did not screen as
potentially risk significant due to external events. (Section 4OA2a(3)(a))
Green. A self-revealing, non-cited violation of 10 CFR 50, Appendix B, Criterion XVI,
Corrective Action, was identified, in that, Entergy failed to adequately evaluate leakage
into the 22 steam generator. During the Indian Point Unit 2 reactor trip on
August 23, 2006, main feedwater low flow bypass valve FCV-427L leaked excessively
and resulted in an uncontrolled rise in 22 steam generator level; operator response to
isolate feedwater to the steam generator in accordance with emergency operating
procedures; and automatic actuation of the feedwater isolation system. The excessive
leakage condition into the 22 steam generator was identified on April 4, 2006, prior to
Indian Point Unit 2 refueling outage 2R17, but was not fully evaluated or corrected prior
to the reactor trip on August 23, 2006. This issue was entered into the corrective action
program, and FCV-427L was repaired and retested satisfactorily.
The inspectors determined that this finding was more than minor because it was
associated with the Equipment Performance attribute of the Mitigating Systems
cornerstone; and, it affected the cornerstone objective of ensuring the availability,
reliability and capability of systems that respond to initiating events to prevent
undesirable consequences. The inspectors evaluated the significance of the finding
Enclosure
iv
using Phase 1 of IMC 0609, Appendix A, "Significance Determination of Reactor
Inspection Findings for At-Power Situations," and determined that the finding was of
very low safety significance because it was not a design or qualification deficiency; it did
not result in the loss of a system safety function or a train safety function for greater
than the Technical Specification Allowed Outage Time; and it did not screen as
potentially risk significant due to external events.
The inspectors determined that the finding had a cross-cutting aspect in the area of
problem identification and resolution because Entergy did not thoroughly evaluate the
cause of excessive leakage into the 22 steam generator such that the resolutions
addressed the causes and extent of condition of the problem. (Section 4OA2a(3)(b))
Cornerstone: Not Applicable
Green. The NRC inspectors identified a finding when Entergy failed to initiate condition
reports in accordance with EN-LI-102, Corrective Action Process, for the adverse
conditions identified in the 2006 Safety Culture Assessment. Consequently, the adverse
conditions were not evaluated and appropriate corrective actions were not identified in a
timely manner. The contractor who performed the independent safety culture
assessment presented the site specific results to Entergy management in June 2006.
The negative responses and declining trends identified in the assessment constituted
adverse conditions that should have been entered into the corrective action program. At
the time of the inspection, Entergy had not initiated condition reports for the assessment
results. Consequently, the results had not been fully evaluated to understand the
causes and identify appropriate actions to address the identified issues. Additionally,
organizations identified by the contractor as needing management attention had not
developed departmental action plans at the time of the inspection. Entergy entered this
issue into the corrective action program and initiated a learning organization condition
report to track development and implementation of action plans to address the
assessment results.
The inspectors determined that the finding was more than minor because if left
uncorrected it would become a more significant safety concern. Without appropriate
action, the weaknesses in the safety culture onsite would continue, increasing the
potential that safety issues would not receive the attention warranted by their
significance. The finding was not suitable for SDP evaluation, but has been reviewed by
NRC management and has been determined to be a finding of very low safety
significance. The finding was not greater than very low safety significance because the
inspectors did not identify any issues that were not raised which had an actual impact on
plant safety or were of more than minor safety significance.
The inspectors determined that this finding had a cross-cutting aspect in the area of
problem identification and resolution because Entergy did not identify issues with the
potential to impact nuclear safety in the corrective action process for evaluation and
resolution in a timely manner. (Section 4OA2c(3))
b.
Licensee-Identified Violations
None.
Enclosure
REPORT DETAILS
4.
OTHER ACTIVITIES (OA)
4OA2 Problem Identification and Resolution (PI&R) (Biennial - IP 71152B)
a.
Assessment of the Corrective Action Program
(1)
Inspection Scope
The inspection team reviewed the procedures describing the Entergy corrective action
program (CAP). Indian Point Unit 2 identified problems for evaluation and resolution by
initiating condition reports (CRs) in the Paperless Condition Reporting System (PCRS).
The team evaluated the methods for assigning and tracking issues to assure that issues
were screened for operability and reportability, prioritized for evaluation and resolution in
a timely manner commensurate with their safety significance, and tracked to identify
adverse trends and repetitive issues. In addition, the team interviewed plant staff and
management to determine their understanding of and involvement with the corrective
action program. The condition reports and other documents reviewed, as well as key
personnel contacted, are listed in the Attachment to this report.
The team reviewed condition reports selected across the seven cornerstones of safety
in the NRCs Reactor Oversight Program (ROP) to determine if problems were being
properly identified, characterized, and entered into the corrective action program for
evaluation and resolution. The team selected items from the maintenance, operations,
engineering, emergency preparedness, physical security, radiation protection, and
oversight programs to ensure that Entergy was appropriately addressing problems
identified in each functional area. The team selected a risk-informed sample of
condition reports that had been issued since the last NRC problem identification and
resolution inspection, which was conducted in June 2005.
The team selected items from other processes at Indian Point to verify that they were
appropriately considered for entry into the corrective action program. Specifically, the
team reviewed a sample of engineering requests (ERs), operability determinations,
maintenance work orders (WOs), engineering system health reports, and completed
surveillance tests. The team also reviewed completed work packages to determine if
issues identified during the performance of preventive maintenance were entered into
the corrective action program. In addition, the team attended operations shift turnover
meetings and accompanied auxiliary operators during rounds in the plant.
The team considered risk insights from both the NRCs and Entergy's risk assessments
for Indian Point Unit 2 to focus the sample selection and plant tours on risk-significant
components. The team determined that the systems with the highest risk significance
were 480V AC, 125V DC, component cooling water, service water, reactor protection,
emergency core cooling system recirculation, safety injection accumulators, and
auxiliary feedwater (AFW). Inspector samples focused on these systems, but were not
limited to them. The review was expanded to five years for evaluation of selected check
valves in the auxiliary feedwater, safety injection and residual heat removal systems.
2
Enclosure
The inspection team reviewed condition reports to assess whether Entergy adequately
evaluated and prioritized identified problems. The issues reviewed encompassed the
full range of evaluations, including root cause analyses, apparent cause evaluations,
and common cause analyses. The review included the appropriateness of the assigned
significance, the scope and depth of the causal analysis, and the timeliness of
resolution. The team observed meetings of the Condition Review Group (CRG), in
which Entergy personnel reviewed new condition reports for prioritization and
assignment, and the Corrective Action Review Board (CARB) where Entergy personnel
evaluated root cause evaluations, as well as selected apparent cause evaluations and
corrective action assignments. The team also reviewed equipment operability
determinations, reportability assessments, and extent-of-condition reviews for selected
problems.
The team reviewed the corrective actions associated with selected condition reports to
determine whether the actions addressed the identified causes of the problems. The
team reviewed condition reports for repetitive problems to determine whether previous
corrective actions were effective. The inspectors also reviewed Entergy's timeliness in
implementing corrective actions and their effectiveness in precluding recurrence for
significant conditions adverse to quality. The team also reviewed condition reports
associated with selected NCVs and findings to determine whether Entergy properly
evaluated and resolved the issues.
(2)
Assessment
Identification of Issues
One Green NCV was identified in the area of identification of issues for failure to identify
improper internal valve clearances on an auxiliary feedwater check valve in the
corrective action program for evaluation and resolution.
In general, the team considered the identification of problems at Indian Point to be
appropriate. The computer-based condition reporting process, PCRS, facilitates the
initiation, tracking and trending of condition reports. Approximately 6,500 condition
reports were written each year. There was a low threshold for the identification of
issues and, in most cases, problems identified during plant activities were entered into
PCRS when appropriate. However, the team found that problems identified in 1997 and
2002 during internal inspections of an auxiliary feedwater check valve were not entered
into the corrective action program for evaluation and resolution. As a result, upon
inspection in September 2006, the gasket between the valve's body to bonnet seal was
found over-crushed causing the gasket to partially unwind, potentially impacting valve
operation. This finding is discussed in detail in Section 4OA2a(3)(a).
Prioritization and Evaluation of Issues
One Green NCV was identified in the area of prioritization and evaluation of issues for
an inadequate evaluation of leakage into the 22 steam generator.
3
Enclosure
The team determined that, in general, Entergy appropriately prioritized and evaluated
issues commensurate with the safety significance of the issue. Condition reports were
screened for operability and reportability, categorized by significance (A through D), and
assigned to a department for evaluation and resolution. The Condition Review Group
appropriately considered human performance issues, radiological safety concerns,
repetitiveness, and adverse trends in their reviews. There were no operability or
reportability determinations with which the team disagreed. However, the team did
identify a condition report that was improperly categorized which led to insufficient
evaluation of the issue. Specifically, the inspectors identified that a condition report
which documented a concern regarding security guard readiness was categorized as a
'D' track and trend CR and closed without evaluation. Following discussions with the
inspectors, Entergy wrote a new condition report to evaluate and address the issue.
Although the failure to evaluate the condition when it was first raised in 2003 does not
comply with NRC requirements, the inspectors determined that due to the nature of the
issue it constitutes a violation of minor significance that is not subject to enforcement
action in accordance with the NRCs Enforcement Policy.
The inspectors found that causal analyses were thorough and appropriately considered
extent of condition, generic issues, and previous occurrences. The Corrective Action
Review Board reviews were detailed and ensured that corrective actions addressed the
identified causes. For significant conditions adverse to quality, corrective actions were
identified to prevent recurrence. However, in one case, the inspectors found that
Entergy did not thoroughly evaluate the cause of excessive leakage into the 22 steam
generator such that the resolution addressed the causes and extent of condition of the
problem which adversely impacted the operators ability to control steam generator water
level following a reactor trip on August 23, 2006. This finding is discussed in detail in
Section 4OA2a(3)(b).
Entergy reviews condition reports site-wide and at the department level to identify
adverse conditions occurring at an unacceptable rate or changes in the frequency or
severity of events or precursors. The team determined that the monthly reviews and
quarterly trend reports provided an effective method for identifying adverse or emerging
trends so that actions could be taken in a timely manner to address the issues.
However, the team identified that some departments did not include 'D' condition reports
in the trending process. Although the 'D' condition reports were included in the site-wide
reviews performed by the Corrective Action and Assessment (CA&A) department, and
system engineers tracked all CRs for their assigned systems, adverse or emerging
trends within a department could have been missed without trending the 'D' condition
reports. Failure to track 'D' condition reports does not comply with Entergy procedure
EN-LI-121, "Entergy Trending Process," but the inspectors did not identify any adverse
or emerging trends that were not identified. Therefore, the inspectors determined that
the issue constitutes a violation of minor significance that is not subject to enforcement
action in accordance with the NRC's Enforcement Policy.
Although the evaluation of issues and determination of required corrective actions was
generally good, the team identified examples of condition descriptions, dispositions
(causal evaluations), and corrective action responses that did not provide clear and
4
Enclosure
complete documentation of the issues and actions taken. This issue had also been
identified by Entergy and they were taking corrective actions to improve the stand
alone quality of CRs. In the identified cases, the inspectors were able to gather
additional information to support the CR packages or the licensee had self-identified the
specific issues in the CAP for resolution.
Effectiveness of Corrective Actions
No findings of significance were identified in the area of effectiveness of corrective
actions.
The team concluded that identified corrective actions were generally appropriate to
resolve identified issues, and were typically completed in a timely manner. The
inspectors also noted a decreasing trend in the number of items in the backlog of
actions to be completed by engineering. However, the inspectors identified a few
instances of incomplete or inadequate corrective actions. For example, on
October 5, 2006, NRC inspectors noted that the positioner feedback arm for Indian
Point Unit 2 high pressure steam dump valve PCV-1121 was not attached to the valve.
Industry operating experience information from 1993 and 1997 identified the need to
incorporate the verification of tightness of valve positioner feedback arms in preventive
maintenance programs due to several incidents caused by feedback arms falling off.
The Indian Point Unit 2 condition report response to this operating experience indicated
that planned maintenance was performed to verify tightness and that mechanisms were
"double nutted" to ensure tightness. NRC inspectors identified that planned
maintenance was not accomplished to verify tightness of positioner feedback arms and
that many of the positioner arms on similar valves were not double nutted. A condition
report was written to address the incomplete corrective actions in response to this
operating experience information. Because the failure of PCV-1121 would not be risk
significant, this issue was determined to be of minor significance.
In the second quarter 2006, the NRC identified several procedure adequacy findings as
documented in IR 50-247/2006-003. In partial response to these findings, Entergy
issued CR IP2-2006-3930 to evaluate the concern and determine appropriate corrective
actions. Entergy efforts to fully scope, prioritize, establish a timeline and take actions to
improve the bulk of operations procedures do not appear to be timely. At the time of
this inspection, Entergy's plan for standardizing operations procedures between the two
units had not been finalized. The completion dates for standardization of the plant
operating procedures appeared to be reasonable; however, the scope and timeline for
the bulk of operations procedures (system operating procedures (SOPs), alarm
response procedures (ARPs), etc.) were not yet defined and preliminarily were targeted
to be completed in calendar year 2010 if the upgrades to SOPs, abnormal operating
procedures, and ARPs were performed in series.
The NRC continued to identify procedure adequacy issues such as incorrect acceptance
criteria in the surveillance procedure for the 22 auxiliary boiler feed pump overspeed trip
test. Specifically, following failure of the auxiliary boiler feed pump during an overspeed
trip test conducted in 2002, the licensee conducted an operability evaluation and
5
Enclosure
determined that the pump was operable and the surveillance acceptance criteria should
be changed. The pump failed the test again in 2006, because timely corrective action
was not taken to revise the surveillance criteria. This issue was of minor significance
because it did not impact the ability of the auxiliary boiler feed pump to perform its safety
function.
The team also identified instances of improper closure of corrective actions to other
processes. Condition reports IP2-2006-0876, IP2-2006-1134, and IP2-2006-2120 were
written to document radiation monitor calibration procedure deficiencies. Corrective
actions for these deficiencies were improperly characterized as enhancements and were
closed out to the procedure feedback process. Closure to this process is not allowed
per EN-LI-102, "Corrective Action Program," unless the change is an enhancement.
Some of the changes to the procedures were required to complete the calibrations;
therefore, the team did not consider the changes to be enhancements. When the
procedures were performed, the procedure deficiencies were handled in accordance
with site administrative procedures; therefore, this issue constitutes a violation of minor
significance that is not subject to enforcement action in accordance with the NRCs
Enforcement Policy. Entergy initiated a condition report to address this issue.
(3)
Findings
(a)
Failure to Identify a Degraded Condition of an Auxiliary Feed Water Check Valve in the
Corrective Action Program
Introduction: The inspectors identified a Green NCV of 10 CFR 50, Appendix B,
Criterion XVI, Corrective Action, in that, Entergy failed to identify a condition adverse to
quality associated with improper internal clearances on BFD-68, an auxiliary feedwater
check valve, in the corrective action program.
Description: On September 1, 2006, Entergy conducted a quarterly test of the 22
auxiliary boiler feed pump (ABFP). During the test, operators observed that an
ultrasonic flow meter indicated no cooling water flow to the pump bearing. The cooling
water flow indication ramped to a normal reading of approximately 32 gallons per minute
after lift check valve BFD-68 was mechanically agitated by the operator. Following the
test, Entergy conducted an internal inspection of the check valve and found that the
spiral-wound gasket for the valves body to bonnet seal had been over-crushed. This
resulted in the gasket becoming partially unwound with a portion of the gasket material
inside the valve cage area, potentially impacting valve operation. It was identified that
the clearance between the valve body and bonnet sealing area was not sufficient to
allow proper gasket crush. Entergy corrected the condition by machining the valve
bonnet to ensure the clearance was appropriate to allow proper gasket crush.
The inspectors observed the field work and reviewed the apparent cause analysis
conducted by Entergy. This evaluation determined the gasket material which intruded
into the valve cage would not likely have prevented valve operation based on the
internal clearances between the valve piston and cage. In addition, no marks were
identified on the internal components which would be indicative of valve misalignment.
6
Enclosure
Based on this, Entergy concluded that the most likely cause of the no flow indication
was an intermittent failure of the ultrasonic flow meter.
The inspectors reviewed the work history associated with BFD-68 and noted that gasket
damage was identified during internal valve inspections performed in 1997 and 2002. In
addition, measurements were taken on the valve body and bonnet during the work in
1997 which indicated the internal clearances were not acceptable. Notes were placed in
the work order packages identifying the gasket damage; however, these deficiencies
were not entered into the licensees corrective action program. Consequently, the
condition was not evaluated and corrected prior to September 2006.
Analysis: The inspectors determined that Entergys failure to identify this degraded
condition and place it in their corrective action program was a performance deficiency, in
that, the improper clearance was a condition adverse to quality that had the potential to
impact operation of a safety-related component. It was reasonable that Entergy should
have identified this deficiency in the corrective action program since the degraded
condition was found during work on the valve and noted in the associated work
packages. Traditional enforcement did not apply since there were no actual safety
consequences or potential for impacting the NRCs regulatory function, and the finding
was not the result of any willful violation of NRC requirements or Entergys procedures.
The inspectors determined that this finding was more than minor because it was
associated with the Equipment Performance attribute of the Mitigating Systems
cornerstone; and, it affected the cornerstone objective of ensuring the availability,
reliability and capability of systems that respond to initiating events to prevent
undesirable consequences. Specifically, the 22 ABFP required approximately three
hours of unplanned unavailability time to conduct repairs to ensure the correct gasket
crush when the valve was reassembled. The inspectors evaluated the significance of
this finding using Phase 1 of IMC 0609, Appendix A, Significance Determination of
Reactor Inspection Findings for At-Power Situations, and determined that the finding
was of very low safety significance (Green) because it was not a design or qualification
deficiency; it did not result in the loss of a system safety function or a train safety
function greater than the Technical Specification Allowed Outage Time; and it did not
screen as potentially risk significant due to external events.
Enforcement: 10 CFR 50 Appendix B, Criterion XVI, Corrective Action, states, in part,
that measures shall be established to assure that conditions adverse to quality, such as
failures, malfunctions, deficiencies, deviations, defective material and equipment and
nonconformances are promptly identified and corrected. Contrary to this, in 1997 and
2002, Entergy failed to identify the improper internal clearances on valve BFD-68 in their
corrective action program. Consequently, the condition was not evaluated and
corrected prior to reassembly of the valve following maintenance in 1997 and 2002.
Entergy subsequently entered this issue into the CAP (CR-IP2-2006-05241), evaluated
the condition, and conducted repairs to the valve to ensure the proper gasket crush was
obtained. Because this issue was of very low safety significance and was entered into
the licensees corrective action program, this violation is being treated as an NCV,
consistent with Section VI.A.1 of the NRC Enforcement Policy: NCV
7
Enclosure
05000247/2006006-02, Failure to Identify a Degraded Condition of an Auxiliary Feed
Water Check Valve in the Corrective Action Program.
(b)
Inadequate Evaluation of Leaking 22 Steam Generator Low Flow Bypass Valve FCV-
427L
Introduction: A Green, self-revealing, non-cited violation of 10 CFR 50 Appendix B,
Criterion XVI, Corrective Action, was identified when Entergy failed to adequately
evaluate leakage into the 22 steam generator. The potential that main feedwater low
flow bypass valve FCV-427L was leaking was identified on April 4, 2006, prior to the
Indian Point Unit 2 refueling outage, but was not fully evaluated or corrected prior to a
reactor trip on August 23, 2006. During the reactor trip on August 23, 2006, FCV-427L
leaked excessively and resulted in actuation of the feedwater isolation system on high
water level in the 22 steam generator.
Description: On April 4, 2006, Entergy identified that steam generator level traces
during several reactor trips dating back to November 2004 showed level increasing at a
rate much faster in 22 steam generator than the other steam generators. Because the
auxiliary feedwater flow and generator steaming rates were identical, it was concluded
the main feedwater regulating valve (FCV-427) and/or low flow bypass valve
(FCV-427L) were leaking at greater than their design leakage rates. Entergy decided to
address potential leakage across FCV-427 during the refueling outage in April and May
2006, since the valve was already planned for overhaul at that time, and to further
evaluate FCV-427L with diagnostic testing in the normal work schedule with the plant
on-line following the refueling outage.
During shutdown on April 19, 2006, for the refueling outage, a large level perturbation
was again noted on the 22 steam generator. Inspection of valve FCV-427 during the
overhaul identified galling, and a blue check revealed that the valve was not seating
properly. Leakage past FCV-427L was not evaluated further at this time and diagnostic
testing of the valve was planned for November 2006, in conjunction with planned
maintenance.
On August 23, 2006, the Indian Point Unit 2 reactor was manually tripped during a plant
transient. Approximately ten minutes after the reactor trip, 22 steam generator level was
noted to be rising rapidly. Operators evaluated the condition, isolated the main and
bypass feedwater valves due to apparent leak-by, and entered Technical Specification 3.7.3 Conditions A.1 and B.1, for FCV-427 and FCV-427L being inoperable. Due to the
rapid increase in steam generator level, a high-high steam generator level signal was
received for the 22 steam generator, resulting in an automatic main feedwater isolation.
An 8-Hour non-emergency event notification report was made per 10 CFR 50.72(b)(3)(iv)(A).
FCV-427 and FCV-427L were inspected during the forced outage and it was identified
that the leakage was from the low flow bypass valve (FCV-427L). The valve stem
stroke was adjusted for FCV-427L and the valve was checked for seat leakage
satisfactorily during the subsequent startup.
8
Enclosure
Analysis: The inspectors determined that Entergys failure to fully evaluate and correct
the excessive leakage into the 22 steam generator was a performance deficiency, in
that, the leakage past the main feedwater low flow bypass valve FCV-427L was a
condition adverse to quality that impacted the ability of the operators to maintain steam
generator water level. It was reasonable that Entergy should have fully evaluated the
source of the leakage during the refueling outage in April and May 2006 prior to the
plant trip in August 2006. Traditional enforcement did not apply since there were no
actual safety consequences or potential for impacting the NRCs regulatory function,
and the finding was not the result of any willful violation of NRC requirements or
Entergys procedures.
The inspectors determined that this finding was more than minor because it was
associated with the Equipment Performance attribute of the Mitigating Systems
cornerstone; and, it affected the cornerstone objective of ensuring the availability,
reliability and capability of systems that respond to initiating events to prevent
undesirable consequences. The leaking low flow bypass valve forced operators to
evaluate and respond to the rapidly rising 22 steam generator level by taking actions to
isolate feed streams to the steam generator during a reactor trip response, and
culminated in an automatic actuation of the feedwater isolation system. The inspectors
evaluated the significance of this finding using Phase 1 of IMC 0609, Appendix A,
Significance Determination of Reactor Inspection Findings for At-Power Situations, and
determined that the finding was of very low safety significance (Green) because it was
not a design or qualification deficiency; it did not result in the loss of a system safety
function or a train safety function greater than the Technical Specification Allowed
Outage Time; and it did not screen as potentially risk significant due to external events.
The inspectors determined that the finding had a cross-cutting aspect in the area of
problem identification and resolution because Entergy did not thoroughly evaluate the
cause of excessive leakage into the 22 steam generator such that the resolution
addressed the causes and extent of condition of the problem.
Enforcement: 10 CFR 50, Appendix B, Criterion XVI, "Corrective Action," states, in part,
that measures shall be established to assure that conditions adverse to quality, such as
failures, malfunctions, deficiencies, deviations, defective material and equipment, and
nonconformances are promptly identified and corrected. Contrary to the above,
Entergy did not fully evaluate and correct the cause of excessive leakage into the 22
steam generator in a timely manner which complicated the response to a plant transient.
The valve stem stroke was adjusted for FCV-427L and the valve was checked for seat
leakage satisfactorily during the subsequent startup. Because this issue was of very low
safety significance (Green) and was entered into the licensees corrective action
program (CR-IP2-2006-05082), this violation is being treated as an NCV consistent with
Section VI.A.1 of the NRC Enforcement Policy: NCV 05000247/2006006-03, Inadequate
Evaluation of Leaking 22 Steam Generator Low Flow Bypass Valve FCV-427L.
9
Enclosure
b.
Assessment of the Use of Operating Experience
(1)
Inspection Scope
The team selected a sample of operating experience issues to confirm that Entergy had
evaluated the operating experience information for applicability to Indian Point Unit 2
and had taken appropriate actions, when warranted. Operating experience (OE)
documents were reviewed to ensure that underlying problems associated with the
issues were appropriately considered for resolution via the corrective action process. A
list of the specific documents reviewed is included in the Attachment to this report.
(2)
Assessment
No findings of significance were identified in the area of operating experience.
The inspectors found that operating experience information was appropriately
considered for applicability, and corrective and preventive actions were taken as
needed. Site OE coordinators screened issues from various sources for applicability to
Indian Point Unit 2 and wrote CRs for additional reviews and corrective actions as
necessary. Operating experience information has been integrated into routine activities,
such as pre-job briefs, procedures, and training material. The inspectors noted several
positive examples in which plant personnel considered operating experience information
in addition to material provided by the Operating Experience Program. In a few cases,
the inspectors found that OE-related CRs had been closed without a closure review by
the site OE coordinator.
c.
Assessment of Self-Assessments and Audits
(1)
Inspection Scope
The team reviewed a sample of CA&A audits, including the most recent audit of the
corrective action program, CAP trend reports, Quality Assurance (QA) audits,
departmental self-assessments, and assessments conducted by independent
organizations. A specific list of documents reviewed is included in the attachment to this
report. These reviews were performed to determine if problems identified through these
assessments were entered into the CAP, when appropriate, and whether corrective
actions were initiated to address identified deficiencies. The effectiveness of the audits
and assessments was evaluated by comparing audit and assessment results against
self-revealing and NRC-identified findings and observations made during the inspection.
The team also reviewed the 2006 Nuclear Safety Culture Assessment, dated March
2006. This was a fleet-wide assessment, conducted by an independent contractor in
early 2006. The inspectors reviewed the assessment report and discussed actions
taken and planned with managers and staff. The inspectors also reviewed corporate
assessments and a departmental teamwork assessment that evaluated similar areas in
order to determine if appropriate action had been taken to address identified issues.
10
Enclosure
(2)
Assessment
One Green finding was identified for failure to enter adverse conditions, which were
identified in an independent safety culture assessment, into the corrective action
program for evaluation and appropriate action.
The team observed that, overall, audits and assessments were critical and, in most
cases, appropriate actions were taken to address identified issues. The inspectors
noted that thorough follow-up reviews were conducted by CA&A, the Self Assessment
Review Board (SARB) and corporate offices. In a few cases, the inspectors found that
appropriate corrective actions were not taken for issues identified during assessments.
For example, an area for improvement (AFI) identified during a self-assessment of
lubrication programs was not captured in a CR for evaluation and tracking. Condition
reports for two other AFIs from the lubrication program assessment were closed without
taking the indicated action. In another case, a corrective action for a radiation protection
QA audit finding related to survey results from personnel contamination events was
closed without addressing the issue. In these cases, the inspectors determined that the
failure to complete these corrective actions were of minor significance due to the nature
of the issues. The inspectors also determined that the results of the 2006 Nuclear
Safety Culture Assessment were not entered into the corrective action program; and as
a result, timely action was not taken to evaluate the results and identify appropriate
corrective actions. This finding is discussed in detail in Section 4OA2c(3).
(3)
Findings
Introduction: A Green finding was identified by the NRC inspectors for failure to initiate
condition reports in accordance with EN-LI-102, Corrective Action Process, for
adverse conditions identified by the 2006 Nuclear Safety Culture Assessment.
Consequently, the adverse conditions were not evaluated and appropriate corrective
actions were not identified in a timely manner.
Description: An independent contractor provided the preliminary results of the 2006
Nuclear Safety Culture Assessment to Entergy in April 2006 and presented the
site-specific results to Entergy management in June. In the assessment report, the
contractor made recommendations to address the negative responses and declining
trends, some of which included management attention to reinforce safety conscious
work environment expectations and behaviors site-wide and in a number of specific
organizations at Indian Point. The contractor also recommended actions to assure the
corrective action program was effective, and immediate action for some organizations
based on comparison of the Indian Point results with industry-wide results.
The General Manager of Plant Operations held a meeting with site managers in July to
discuss the results for Indian Point and actions needed. Managers were directed to
discuss the results of the assessment with their staffs, but the results of the assessment
were not entered into the CAP and no specific followup actions were assigned at that
time.
11
Enclosure
At the time of the inspection, an action plan was being developed at the corporate level
which would be customized by each site based on the site specific results and identified
areas for improvement. The draft action plan for the site indicated that an assessment
of safety culture performance indicators from across various disciplines should be done
at Indian Point to understand the causes for the identified issues so that effective
corrective actions could be taken. Entergy management planned to perform another
independent assessment in early 2007 to complete this action. The inspectors
considered this assessment to be untimely, in that, it would not provide insight into the
causes of the issues identified by the 2006 assessment since the data would be
collected more than a year after the original assessment was performed.
The inspectors found that specific actions to address the organizations identified by the
contractor as needing management attention were not initiated until mid-September.
The draft corporate action plan also indicated that departmental action plans would be
developed for these organizations based on review of the department-specific safety
culture assessment results in conjunction with an assessment of the department's safety
culture performance indicators. At the time of the inspection, most of these
organizations had only recently reviewed the department-specific safety culture
assessment results and were in the early stages of developing department action plans.
Entergy Procedure EN-LI-102, Corrective Action Program, requires the initiation of
condition reports for adverse conditions, which are defined as conditions that detract
from safe nuclear plant operation or that could credibly impact nuclear safety. The
inspectors concluded that the negative responses and declining trends identified by the
safety culture assessment could impact nuclear safety because the assessment results
were precursors and indicators of a possible reluctance to raise safety issues by site
employees, particularly in certain organizations. Therefore, these results should have
been considered adverse conditions that warranted initiation of a condition report.
Failure to enter the assessment results into the corrective action program resulted in a
delay in evaluating the results to understand the causes and identify appropriate
corrective or mitigative actions.
Analysis: The inspectors determined that Entergy's failure to enter the adverse
conditions identified during the 2006 Nuclear Safety Culture Assessment into the
corrective action program and evaluate the results and identify appropriate corrective
actions in a timely manner was a performance deficiency that was reasonably within
Entergy's ability to foresee and correct. Traditional enforcement did not apply since
there were no actual safety consequences or potential for impacting the NRC's
regulatory function, and the finding was not the result of any willful violation of NRC
requirements or Entergy's procedures.
The inspectors determined that this finding was more than minor because if left
uncorrected it would become a more significant safety concern. Without appropriate
action, the weaknesses in the safety culture onsite would continue, increasing the
potential that safety issues would not receive the attention warranted by their
significance. The finding was not suitable for SDP evaluation, but has been reviewed by
NRC management and has been determined to be a finding of very low safety
12
Enclosure
significance (Green). The finding was not greater than very low safety significance
because the inspectors did not identify any issues that were not raised which had an
actual impact on plant safety or were of more than minor safety significance.
The inspectors determined that this finding had a cross-cutting aspect in the area of
problem identification and resolution because Entergy did not identify adverse conditions
with the potential to impact nuclear safety in the corrective action process for evaluation
and resolution in a timely manner.
Enforcement: No violation of NRC regulatory requirements was identified. Although
Entergy did not initiate condition reports for the adverse conditions identified by the
safety culture survey, application of EN-LI-102 for these conditions does not fall under
NRC regulatory requirements. After identification by the team, Entergy entered this
issue into the CAP (CR IP2-2006-06105) and initiated a Learning Organization (LO)
condition report to track development and implementation of site and department action
plans to address the assessment results. Because this finding does not involve a
violation of regulatory requirements and has very low safety significance, it is identified
as FIN 05000247/2006006-01, Failure to Enter Safety Culture Assessment Results into
Corrective Action Program.
d.
Assessment of Safety Conscious Work Environment
(1)
Inspection Scope
During interviews and discussions with station personnel, the team assessed the safety
conscious work environment (SCWE) at Indian Point. Specifically, the inspectors
assessed whether workers were willing to enter issues into the corrective action
program or raise safety concerns to their management and/or the NRC. The inspectors
conducted individual interviews and held discussions with staff and supervisors
regarding use of the corrective action program, work processes, and other problem
identification and resolution activities. The team reviewed the Indian Point Employee
Concerns Program (ECP) to assess whether employees were willing to use the program
as an alternate path for raising concerns. The team also reviewed a sample of the ECP
files to ensure that issues were appropriately addressed.
(2)
Assessment
No findings of significance were identified.
The team found that most workers indicated that they would raise issues that they
recognized as nuclear safety issues. However, the inspectors also found that a number
of workers interviewed indicated that they were aware of individuals they perceived as
having been treated negatively by management for raising issues; most of these
workers were in the Instrumentation and Controls (I&C) department. Some workers
expressed reluctance to raise issues under certain circumstances due to a number of
reasons, including fear of disciplinary action and concerns with the efficacy of the
corrective action program. While most workers made a distinction between nuclear
13
Enclosure
safety issues and other concerns, the inspectors noted that some of the illustrative
examples provided by plant workers could have nuclear safety implications (i.e.,
procedure quality and staff qualification issues). In one case, a worker indicated that
he/she would not raise issues under any circumstances. In another case, a worker
indicated that he/she had not raised a specific nuclear safety issue. The inspectors
determined that although the issue was a nuclear safety issue, it did not have an actual
impact on safe plant operation in this particular instance due to the specific
circumstances surrounding the issue.
The team determined that the reluctance to raise issues expressed by the I&C staff was
the result of several factors, primarily the fear of disciplinary action compounded by
unclear expectations and standards, and to some extent a lack of confidence in the
corrective action program. The majority of the I&C staff interviewed described instances
which they perceived to be either a negative reaction from management or employee
discipline for raising issues. The inspectors observed that expectations for writing CRs
were not clearly understood within the I&C department which may have contributed to
the perception that individuals were disciplined for raising issues. The inspectors also
found that expectations and standards in other areas, such as qualification and
procedure requirements, were also unclear and contributed to the negative views
expressed by some of the individuals. A number of interviewees also believed that
issues that did not directly impact plant operations, such as personnel or industrial
safety issues, would not be resolved or corrected by the corrective action program.
The team also determined that negative perceptions similar to those in the I&C
department existed in other site organizations. For example, within the Operations
department there was some apprehension about the perceived increase in disciplinary
actions within the department. Additionally, a number of individuals did not have
confidence that the corrective action program would resolve issues of lesser
significance, particularly repeat issues. Based on a limited review, the team found
similar issues, but to a lesser extent, in other departments. Consequently, the team was
concerned that the lack of confidence in the corrective action program and the
apprehension about disciplinary action could challenge the free flow of information and
result in reluctance to raise issues in other departments.
Entergy has self-identified areas for improvement intended to enhance employee
confidence in the corrective action program and the ECP, and has taken actions to
address negative employee perceptions. However, the team determined that these
efforts have not been fully effective in establishing employee confidence in these
programs. For example, the Corrective Action and Assessment department has taken
actions to improve the quality of feedback to employees, but the inspectors found
several examples of corrective action responses that did not provide appropriate
documentation of how the issue was resolved. As described above, the team found that
many employees still have the perception that lower level issues will not be resolved by
the corrective action program. In addition, during interviews, very few workers identified
the ECP as an alternate path for raising issues, and most of those that referenced the
ECP did not view the program as a viable path for raising issues primarily due to
concerns about confidentiality of the program.
14
Enclosure
4OA6 Meetings, including Exit
On December 5, 2006, the team presented the inspection results to Mr. F. Dacimo and
and other Entergy personnel, who acknowledged the findings. The inspection findings
and observations were also discussed with Entergy management during a
teleconference on December 14, 2006. The inspectors confirmed that proprietary
information reviewed during the inspection would be handled in accordance with 10 CFR 2.390, "Public Inspections, Exemptions, Request for Withholding."
ATTACHMENT: Supplemental Information
In addition to the documentation that the inspectors reviewed (listed in the attachment),
copies of information requests given to the licensee are in ADAMS, under accession
number ML063490222.
Attachment
ATTACHMENT - SUPPLEMENTAL INFORMATION
KEY POINTS OF CONTACT
Licensee Personnel
V. Andreozzi, Electrical Systems Manager
J. Balla, Employee Concerns Program Manager
R. Buckley, Corrective Actions Self Assessment Coordinator
R. Burroni, Assistant Operations Manager - Operations Support
V. Cambigianis, Mechanical Design Manager
S. Carpenter, Maintenance Department Corrective Actions Coordinator
J. Comiotes, Director of Nuclear Safety Assurance
J. Conforti, Maintenance Procedure Coordinator
F. Dacimo, Site Vice President
A. Deland, QA Self Assessment & Corrective Actions Coordinator
J. Donnelly, Director of Maintenance
R. Hansler, Reactor Engineering Manager
M. Hornyak, Project Manager, Operations Support ENN, Operating Experience Department
L. Kelly, Planning, Scheduling & Outage Corrective Actions Coordinator
D. Loope, Radiation Protection Manager
S. Meighan, Radiation Protection CA&A Supervisor
E. O'Donnell, Manager - Unit 2 Operations
D. Parker, Maintenance Superintendent
J. Perotta, Quality Assurance Manager
B. Ray, Assistant Superintendent - I&C
P. Rubin, General Manager Plant Operations
A. Small, Manager - Planning, Scheduling and Outage
B. Taggert, Employee Concerns Program Coordinator
M. Tumicki, CA Corrective Actions Coordinator
S. Verrocki, Systems Manager
T. Vitale, Operations Manager
R. Walpole, Corrective Actions Manager
Contractor Personnel
H. Levin, Synergy Consulting Services Corporation
A-2
Attachment
LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED
Opened and Closed
Failure to Enter Safety Culture Assessment Results into
Corrective Action Program (Section 4OA2c(3))05000247/2006006-02
Failure to Identify a Degraded Condition of an Auxiliary
Feed Water Check Valve in the Corrective Action Program
(Section 4OA2.a(3)(a))05000247/2006006-03
Inadequate Evaluation of Leaking 22 Steam Generator
Low Flow Bypass Valve FCV-427L (Section 4OA2.a(3)(b))
A-3
Attachment
LIST OF DOCUMENTS REVIEWED
Procedures and Instructions
0-LUB-401-GEN, "Lubrication of Plant Equipment," Rev 2
0-MD-401, "Management Control of Maintenance Training," Rev 1
0-MD-402, "Maintenance Procedure Development and Feedback Administrative Directive,"
Rev 1
0-VLV-413-MOV, "Motor Operated Valve Minor Preventative Maintenance," Rev 1
2-VLV-012-VCK, "Velan Swing Check Valves," Rev 0
AOV-B-027-A, "Generic Procedure for Testing AOVs Using the MOVATS Diagnostic Test
System, Rev 5
E-0, "Reactor Trip or Safety Injection," Rev 47
E-0, "Reactor Trip or Safety Injection Background Document," Rev 46
E-2, "Faulted Steam Generator Isolation," Rev 39
E-3, "Steam Generator Tube Rupture, Rev 45
ECA-2.1, "Uncontrolled Depressurization of All Steam Generators," Rev 43
EN-DC-118, "Engineering Change Closure," Rev 0
EN-DC-119, "Equipment Database (EDB) Process and Controls," Rev 0
EN-DC-134, "Design Verification," Rev 0
EN-EC-100, Guidelines for Implementation of the Employee Concerns Program, Rev 1
EN-LI-102, Corrective Action Process, Rev 7
EN-LI-104, Self-Assessment and Benchmark Process, Rev 2
EN-LI-118, Root Cause Analysis Process, Rev 4
EN-LI-119, Apparent Cause Evaluations (ACE) Process, Rev 3
EN-LI-121, Entergy Trending Process, Rev 3
EN-MA-123, Identification and Trending of Rework, Rev 0
EN-OE-100, "Operating Experience Program," Rev 2
EN-PL-190, Maintaining a Strong Safety Culture, Rev 0
EN-PL-187, Safety Conscious Work Environment Policy, Rev 0
EN-QV-109, "Audit Process"
EN-RP-104, Personnel Contamination Events, Rev 1
EN-WM-100, Work Request (WR) Generation, Screening and Classification, Rev 0
EN-WM-105, "Planning," Rev 0
ENN-DC-112, Engineering Request and Project Initiation Process, Rev 7
ENN-DC-128, "Calculations," Rev 6
ES-0.1, "Reactor Trip Response," Rev 43
I&C Preventive Maintenance Package No. 1587, MS/HP Steam Dump Valves, Rev 1
IP-SMM-AD-102, "IPEC Implementing Procedure, Preparation, Review, and Approval"
IP-SMM-OP-106, "Procedure Use and Adherence"
IP-SMM-WM-100, "Work Management Process," Rev 4
PT-R99, "HP Steam Dump Stroke Test," Rev 3
VCK-B-021-A, "Generic Procedure for Testing Check Valves Using the MOVATS Diagnostic
Test System," Rev 1
Procedure Feedback Forms
IP3-1320, dated April 26, 2006
A-4
Attachment
IP3-1321, dated April 26, 2006
IP3-1322, dated April 26, 2006
IP3-1355, dated May 8, 2006
IP3-1364, dated June 1, 2006
IP3-1371, dated June 28, 2006
IP3-1383, dated July 5, 2006
Audits and Assessment Reports
QA Audits
QA-03-2005-IP-1, "IPEC Corrective Action Program," May 2005
QA-08-2005-IP-1, "IPEC Unit 3 Engineering Programs"
QA-04-2006-IP-1, "Design Control"
QA-07-2006-IP-1, "IPEC Emergency Planning Audit," May 2006
QA-10-2005-IP-1, "IPEC Maintenance Program," June 2006
QA-12-2005-IP-1, "IPEC Operations Program"
QA-14-2006-IP-1, IPEC Radiation Protection Program, April 2006
QA-16-2005-IP-1, "IPEC Security Audit," March 2006
QA Surveillances
QS-2006-IP-05, "Initial Licensed Operator (ILO) Training Program"
QS-2006-IP-10, "Followup Assessment on Corrective Actions from Security Audit"
QS-2006-IP-15, "Refueling - Fuel Receipt, Core Unload, Core Reload"
Oversight Observation Checklists
O2C-IPEC-2005-0097, "Use of Operating Experience throughout the IPEC Station"
O2C-IPEC-2005-0188, "Inquiry by QA Manager as to how manual CRs and Operability reviews
are being handled by Operations"
O2C-IPEC-2006-0477, "Control Room Observation and AOT Entry"
O2C-IPEC-2006-0685, "Conduct of Operations"
O2C-IPEC-2006-0691, "Control Room Observation and Startup"
O2C-IPEC-2006-0844, "Operations Procedures and Documentation"
O2C-IPEC-2006-0873, "July/August 2006 Monthly SCWE Summary for Site Meeting
Attendance"
Assessments (Learning Organization Condition Reports)
IP3LO-2005-00075, "Conduct of I&C -I&C Department Assessment of Teamwork and Trust
(Snapshot)"
IP3LO-2005-00102, "Compliance with Gun Room Procedures (Snapshot)"
IP3LO-2005-00108, "Line Ownership of CR Trending (Snapshot)"
IP3LO-2005-00119, "Security Officers Knowledge of Human Performance Error Prevention
Tools & Error Traps (Snapshot)"
IP3LO-2005-00147, "Human Performance Self Assessment"
IP3LO-2005-00168, "Fourth Quarter 2005 CA&A Self-Assessment (Ongoing)"
A-5
Attachment
IP3LO-2005-00207, "Problem Identification and Resolution"
IP3LO-2005-00216, "Lubrication/Predictive Maintenance Program (Focused)"
IP3LO-2005-00219, "Human Performance Self Assessment"
IP3LO-2005-00222, "EP Non-siren Equipment Self Assessment (Focused)"
IP3LO-2005-00224, "Performance of Supplemental Personnel"
IP3LO-2005-00298, "Effectiveness of Corrective Action Closures to Lower Tier Monitored
Processes (Snapshot)"
IP3LO-2005-00307, "IPEC Safety Culture Corporate Assessment"
IP3LO-2005-00314, "Anonymous Condition Reports (Ongoing)"
IP3LO-2006-00003, "Conservative Decision Making"
IP3LO-2006-00014, "Cross-cutting Root Cause Issues (Snapshot)"
IP3LO-2006-00016, "Work Package Generation and Distribution"
IP3LO-2006-00038, "Compliance with Gun Room Procedures (Snapshot)"
IP3LO-2006-00072, "Second Quarter 2006 CA&A Self-Assessment (Ongoing)"
IP3LO-2006-00138, "Operations Training Accreditation"
IP3LO-2006-00140, "PI&R Self Assessment (Focused)"
IP3LO-2006-00158, "Corrective Actions, OE, and Human Performance in Emergency Planning
Department (Focused)"
IP3LO-2006-00166, "Periodic Review of RP Standing Orders and RP Standards (Snapshot)"
IP3LO-2006-00206, "Common Cause of Emergent Work in 2005 (Snapshot)"
IP3LO-2006-00331*, "Track Development and Implementation of Action Plans to Improve
Safety Culture in Various IPEC Departments"
"IPEC 2004 Operating Experience Program Self Assessment"
"Indian Point Nuclear Station's Corporate Follow-up Assessment"
"2006 Nuclear Safety Culture Assessment" (Proprietary)
Trend Reports
IPEC Quarterly Trend Report -Third Quarter 2005
IPEC Quarterly Trend Report -Fourth Quarter 2005
IPEC Quarterly Trend Report -First Quarter 2006
IPEC Quarterly Trend Report -Second Quarter 2006
Condition Reports (* denotes a CR generated as a result of this inspection)
Common Issues
IP2-2005-04290
IP2-2005-04475
IP2-2005-05095
IP2-2006-00208
IP2-2006-00676
IP2-2006-00903
IP2-2006-01199
IP2-2006-01202
IP2-2006-02120
IP2-2006-02156
IP2-2006-02747
IP2-2006-02796
IP2-2006-03704
IP2-2006-03881
IP2-2006-04158
IP2-2006-04311
IP2-2006-05105
IP2-2006-05109
IP2-2006-05553
IP2-2006-05968*
A-6
Attachment
IP3-2004-03072
IP3-2004-03774
IP3-2005-00462
IP3-2005-00605
IP3-2005-00723
IP3-2005-01346
IP3-2005-01912
IP3-2005-02624
IP3-2005-03125
IP3-2005-03697
IP3-2005-03881
IP3-2005-04303
IP3-2005-04305
IP3-2005-04307
IP3-2005-04654
IP3-2005-05724
IP3-2006-00006
IP3-2006-00338
IP3-2006-00547
IP3-2006-02112
IP3-2006-02191
IP3-2006-02833
IP3-2006-03099*
IP3-2006-03150*
Unit 2
IP2-1998-05365
IP2-2002-04096
IP2-2002-08794
IP2-2002-10565
IP2-2002-11029
IP2-2003-00111
IP2-2003-01415
IP2-2003-05219
IP2-2003-05608
IP2-2004-00090
IP2-2004-00099
IP2-2004-02447
IP2-2004-04129
IP2-2004-04624
IP2-2004-06526
IP2-2005-00252
IP2-2005-01846
IP2-2005-01975
IP2-2005-02557
IP2-2005-03183
IP2-2005-03245
IP2-2005-03288
IP2-2005-03516
IP2-2005-03555
IP2-2005-03898
IP2-2005-04124
IP2-2005-04178
IP2-2005-04309
IP2-2005-04311
IP2-2005-04412
IP2-2005-04570
IP2-2005-04655
IP2-2005-04926
IP2-2005-05258
IP2-2006-00045
IP2-2006-00111
IP2-2006-00203
IP2-2006-00421
IP2-2006-00551
IP2-2006-00553
IP2-2006-00647
IP2-2006-00732
IP2-2006-01011
IP2-2006-01033
IP2-2006-01134
IP2-2006-01664
IP2-2006-01834
IP2-2006-01897
IP2-2006-01967
IP2-2006-01988
IP2-2006-02023
IP2-2006-02025
IP2-2006-02058
IP2-2006-02109
IP2-2006-02136
IP2-2006-02168
IP2-2006-02175
A-7
Attachment
IP2-2006-02217
IP2-2006-02221
IP2-2006-02233
IP2-2006-02259
IP2-2006-02313
IP2-2006-02778
IP2-2006-03122
IP2-2006-03199
IP2-2006-03242
IP2-2006-03260
IP2-2006-03354
IP2-2006-03502
IP2-2006-03681
IP2-2006-03707
IP2-2006-03837
IP2-2006-03903
IP2-2006-03931
IP2-2006-04017
IP2-2006-04081
IP2-2006-04270
IP2-2006-04282
IP2-2006-04311
IP2-2006-04412
IP2-2006-04552
IP2-2006-04672
IP2-2006-04802
IP2-2006-04858
IP2-2006-04862
IP2-2006-05013
IP2-2006-05036
IP2-2006-05082
IP2-2006-05114
IP2-2006-05201
IP2-2006-05241
IP2-2006-05453
IP2-2006-05530*
IP2-2006-05532*
IP2-2006-05568*
IP2-2006-05872*
IP2-2006-05897*
IP2-2006-05901*
IP2-2006-06114*
Operating Experience Reviews
EA 03-09, "Head and nozzle inspection"
IN 1998-018, "Recent Contamination Incidences Resulting from Failure to Perform Adequate
Surveys"
IN 2004-019, "Problems Associated with Back-up Power Supplies to Emergency Response
Facilities and Equipment"
IN 2005-006, "Failure to Maintain Alert and Notification System Alert Radio Capability"
IN 2005-015, "PVNGS 3 Unit Trip"
IN 2005-024, "RCS Leak Detection"
IN 2005-030, "Safe Shutdown Potentially Challenged by Unanalyzed Internal Flooding Events
and Inadequate Design"
IN 2006-009, "Performance of NRC-Licensed Individuals While on Duty with Respect to Control
Room Attentiveness"
LO-OEN-2003-0364, "Reactor trip due to decreasing vacuum"
OE 2378, "MOV Motor Bolting Failure"
OE 15162, "Containment Spray (CS) Valve 1CS001A Failed to Open During Quarterly
A-8
Attachment
Maintenance Work Orders
00-18584
01-19922
02-31553
02-64068
03-03458
03-19925
04-19881
05-12018
05-12019
05-19804
05-23762
06-00079
06-21098
06-27618
06-27635
06-27637
06-28849
Work Requests
IP2-04-33937
Non-Cited Violations and Findings Reviewed
FIN 50-247/2005-03-02
Inadequate corrective actions associated with training, procedural
adequacy and operator knowledge on methods to address
degraded grid
NCV 50-247/2005-05-04
Inadequate procedure from control of work on safety-related
components
NCV 50-247/2005-05-05
Inadequate equipment to assess threshold for EAL 8.4.3
FIN 50-247/2005-05-06
Inadequate Corrective Actions for Frame Relay System Problems
NCV 50-247/2005-05-07
Failure to make 10 CFR 50.72 notification for siren problems
FIN 50-247/2005-08-01
Inadequate surveillance testing of TSC diesel generator
NCV 50-247/2006-02-02
Failure to effectively control the performance of the rod position
indication system
NCV 50-247/2006-02-04
Scaffolding control issue results in reactor trip
NCV 50-247/2006-03-01
Inadequate procedure for placing RHR pump suction gauges in
service
NCV 50-247/2006-03-05
Inadequate post-work test on 21 EDG
NCV 50-247/2006-03-06
Inadequate procedure for venting the reactor vessel head while
shut down
NCV 50-247/2006-03-07
Failure to assess the risk of maintenance activities on valve SI-
869A
NCV 50-247/2006-03-09
Failure to implement procedure requirements associated with core
support barrel replacement
NCV 50-247/2006-03-10
Failure to perform adequate surveys to evaluate radiation levels
during core support barrel replacement
FIN 50-247/2006-03-11
Inadequate procedure for placing standby main lube oil cooler in
service
System Health Reports
2nd Quarter 2006 Auxiliary Feed Water System
Miscellaneous
A-9
Attachment
Entergy memorandum dated June 3, 2006, "Expectations for Condition Report Initiation," from
F. Dacimo, Site Vice President, to IPEC Managers
Entergy memorandum IPEC-ADM-06-008, dated February 8, 2006, "Expectations," from
P. Rubin, General Manager Plant Operations, to Managers and Supervisors
Entergy Nuclear Northeast Operating Experience Program Monthly Reports for July and August
2006
Indian Point Energy Center 2006 Second Quarter Report
IPEC Maintenance Rule Basis Document, Main Feedwater System, Units 2&3, Rev 0
IPEC Project/Team Lead Job Familiarization & Professional Development Guide, Rev 1
Inside Entergy Tailgate Edition dated September 21, 2006
Inside Entergy dated April 24, 2006, "Preliminary Results of the Synergy Nuclear Safety Culture
Assessment (NSCA)"
Inside Entergy dated August 6, 2006, "Heron, Campbell Address Nuclear Safety Culture
Results"
Morale Committee Newsletters dated September 2005, March 2006 and July 2006
Root Cause Analysis: "Worker Exceeded Radiological Administrative Setpoint During Lower
Internals Move," dated June 1, 2006
Tailgate article dated August 17, 2006, "PCRS Operability and Immediate Reportability
Screening"
Tailgate article dated August 17, 2006, "The Condition Reporting Process"
Tailgate article dated August 24, 2006, "The Importance of Stand-Alone Quality"
Tailgate article dated August 31, 2006, "Condition Report Initiation"
Tailgate article dated September 7, 2006, "What Makes a Good Condition Report?"
Talking Points for GMPO Web Page, dated October 4, 2006
Ultrasonic Testing Plan, IPEC Utility Tunnel, Rev 0
Westinghouse Technical Bulletin TB-04-22, Reactor Coolant Pump Seal Performance -
Appendix R Compliance and Loss of All Seal Cooling, Rev 1
A-10
Attachment
LIST OF ACRONYMS
ABFP
Auxiliary Boiler Feed Pump
Apparent Cause Evaluation
Agency Document Administrative Management System
Area for Improvement
Auxiliary Feedwater System
Alarm Response Procedure
Corrective Action and Assessment
Corrective Action Program
Corrective Action Review Board
CFR
Code of Federal Regulations
CR
Condition Report
Condition Review Group
Direct Current
Division of Reactor Projects
Employee Concerns Program
ER
Engineering Request
Flow Control Valve
Finding
Instrumentation and Controls
IMC
NRC Inspection Manual Chapter
IN
NRC Information Notice
IP
NRC Inspection Procedure
Indian Point Energy Center
IR
NRC Inspection Report
LCO
Limiting Condition for Operation
Non-Cited Violation
NRC
Nuclear Regulatory Commission
Operating Experience
Publicly Available Records
Paperless Condition Reporting System
Problem Identification & Resolution
Quality Assurance
Residual Heat Removal System
Reactor Oversight Program
Self-Assessment Review Board
Safety Conscious Work Environment
Significance Determination Process
System Operating Procedure
Work Order