ML111400351
| ML111400351 | |
| Person / Time | |
|---|---|
| Site: | Wolf Creek  |
| Issue date: | 05/20/2011 |
| From: | Geoffrey Miller NRC Region 4 |
| To: | Matthew Sunseri Wolf Creek |
| References | |
| IR-11-006 | |
| Download: ML111400351 (34) | |
See also: IR 05000482/2011006
Text
UNITED STATES
NUCLEAR REGULATORY COMMISSION
REGI ON I V
612 EAST LAMAR BLVD, SUITE 400
ARLINGTON, TEXAS 76011-4125
May 20, 2011
Matthew W. Sunseri, President and
Chief Executive Officer
Wolf Creek Nuclear Operating Corporation
P.O. Box 411
Burlington, KS 66839
SUBJECT: WOLF CREEK GENERATING STATION - NRC INSPECTION
PROCEDURE 95002 SUPPLEMENTAL INSPECTION REPORT AND
ASSESSMENT FOLLOWUP LETTER 05000482/2011006
Dear Mr. Sunseri:
On March 31, 2011, the U.S. Nuclear Regulatory Commission (NRC) completed a supplemental
inspection pursuant to Inspection Procedure 95002, "Supplemental Inspection for One
Degraded Cornerstone or Any Three White Inputs in a Strategic Performance Area," at your
Wolf Creek Generating Station facility. The supplemental inspection also covered the
performance issues associated with Inspection Procedure 92723, "Follow Up Inspection for
Three or More Severity Level IV Traditional Enforcement Violations in the Same Area in a
12-Month Period. The enclosed inspection report documents the inspection results, which
were discussed at the exit meeting on April 5, 2011, with yourself and other members of your
staff.
As required by the NRC Reactor Oversight Process Action Matrix, this supplemental inspection
was performed to address three white performance indicators associated with unplanned
scrams, unplanned scrams with complications, and safety system functional failures. These
performance issues were documented previously on the NRC public web page
(http://www.nrc.gov/NRR/OVERSIGHT/ASSESS/WC/wc_chart.html). The NRC staff was
informed on October 29, 2010, of your staff's readiness for this inspection.
The objectives of this supplemental inspection were to: 1) provide assurance that the root
causes and the contributing causes for the risk-significant performance issues were understood;
2) provide assurance that the extent-of-condition and extent-of-cause of the issues were
identified; and 3) provide assurance that corrective actions were sufficient to address and
prevent the recurrence of the root and contributing causes. This inspection also included
independent NRC reviews of the extent-of-condition and extent-of-cause for the three white
performance indicators and assessments of whether any safety culture component caused or
significantly contributed to the issue. The inspection consisted of examination of activities
conducted under your license as they relate to safety, compliance with the Commission's rules
and regulations, and the conditions of your operating license.
Wolf Creek Nuclear Operating Corporation -2-
The inspectors determined that your staff performed a comprehensive evaluation of individual
and collective causes of the three White performance indicators. Your staff's evaluation
identified root causes of the issues to be: 1) inadequate management oversight/standards
enforcement, 2) lack of knowledge across the station concerning the components of nuclear
safety culture and crosscutting issues, and 3) inadequate hardware monitoring. The inspectors
determined that your staff proposed appropriate corrective actions to upgrade preventative
maintenance practices, improve system health through operating experience reviews, improve
the effectiveness of management review processes, and address deficiencies related to safety
culture which, if successfully implemented, will resolve the identified performance issues. With
respect to Inspection Procedure 92723, the inspectors determined that your staff identified the
causes of the traditional enforcement violations, performed an adequate review of the extent-of-
condition and extent-of-cause, and identified appropriate corrective actions sufficient to address
the causes.
On May 3, 2011, using the results of this inspection, the NRC staff completed a quarterly review
of plant performance of Wolf Creek Generating Station. The assessment also evaluated the
performance indicators and the remaining inspection results for the first quarter of calendar year
2011. We noted that the Safety Systems Functional Failure Performance Indicator returned to
Green at the beginning of the second quarter of 2010. This letter supplements, but does not
supersede, our end-of-cycle assessment letter issued on March 4, 2011.
Overall, Wolf Creek operated in a manner that preserved the publics health and safety and fully
met the cornerstone objectives. All inspection findings for the assessment period were
classified as having very low safety significance (Green) and all performance indicators
indicated performance within the nominal, expected range (Green). As a result, we have
assessed Wolf Creek to be in the Licensee Response column of the NRCs Action Matrix.
Therefore we plan to conduct baseline inspection during the remainder of the assessment cycle.
Based on the results of this inspection, the NRC has identified one issue that was evaluated
under the risk significance determination process as having very low safety significance
(Green). The NRC has also determined that a violation is associated with this issue. Because
of the very low safety significance and because it is entered into your corrective action program,
the NRC staff is treating this finding as a noncited violation consistent with Section 2.3.2 of the
NRC Enforcement Policy. If you contest any noncited violation in this report, you should provide
a response within 30 days of the date of this inspection report, with the basis for your denial, to
the Nuclear Regulatory Commission, ATTN.: Document Control Desk, Washington DC 20555-
0001; with copies to the Regional Administrator, Region IV; the Director, Office of Enforcement,
United States Nuclear Regulatory Commission, Washington, DC 20555-0001; and the NRC
Resident Inspector at the Wolf Creek Generating Station.
In accordance with 10 CFR 2.390 of the NRC's "Rules of Practice," a copy of this letter, its
enclosure, and your response (if any) will be available electronically for public inspection in the
NRC Public Document Room or from the Publicly Available Records component of NRC's
Wolf Creek Nuclear Operating Corporation -3-
document system (ADAMS). ADAMS is accessible from the NRC Website at
http://www.nrc.gov/reading-rm/adams.html (the Public Electronic Reading Room).
Sincerely,
/RA/Rick Deese for
Geoffrey B. Miller, Chief
Project Branch B
Division of Reactor Projects
Docket: 50-482
License: NPF-42
Enclosure:
Inspection Report 05000482/2011006
w/Attachment 1: Supplemental Information
Distribution via ListServ for Wolf Creek Nuclear Operating Corporation
Wolf Creek Nuclear Operating Corporation -4-
Electronic distribution by RIV:
Regional Administrator (Elmo.Collins@nrc.gov)
Deputy Regional Administrator (Art.Howell@nrc.gov)
DRP Director (Kriss.Kennedy@nrc.gov)
DRP Deputy Director (Troy.Pruett@nrc.gov)
DRS Director (Anton.Vegel@nrc.gov)
DRS Deputy Director (Vacant)
Senior Resident Inspector (Chris.Long@nrc.gov)
Resident Inspector (Charles.Peabody@nrc.gov)
WC Administrative Assistant (Shirley.Allen@nrc.gov)
Senior Project Engineer, DRP/B (Rick.Deese@nrc.gov)
Senior Project Engineer, DRP/B (Leonard.Willoughby@nrc.gov)
Public Affairs Officer (Victor.Dricks@nrc.gov)
Public Affairs Officer (Lara.Uselding@nrc.gov)
Project Manager (Randy.Hall@nrc.gov)
Branch Chief, DRS/TSB (Michael.Hay@nrc.gov)
RITS Coordinator (Marisa.Herrera@nrc.gov)
Regional Counsel (Karla.Fuller@nrc.gov)
Congressional Affairs Officer (Jenny.Weil@nrc.gov)
OEMail Resource
ROPreports
Executive Technical Assistant (Stephanie.Bush-Goddard@nrc.gov)
DRS/TSB STA (Dale.Powers@nrc.gov)
R:\ _REACTORS\_WC\2011\ WC2011006 ded-rp.docx
ADAMS: No Yes SUNSI Review Complete Reviewer Initials: GM
Publicly Available Non-Sensitive
Non-publicly Available Sensitive
DDumbacher KKennedy GMiller
RIV:SRI/B D/DRP C/DRPB
/E-DDumbacher for/ /VGG for/ /RA/
5/17/11 5/20/11 5/18/11
OFFICIAL RECORD COPY T=Telephone E=E-mail F=Fax
NUCLEAR REGULATORY COMMISSION
REGION IV
Docket: 05000482
License: NPF-42
Report: 05000482/2011006
Licensee: Wolf Creek Nuclear Operating Corporation
Facility: Wolf Creek Generating Station
Location: 1550 Oxen Lane SE
Burlington, Kansas
Dates: February 7 through March 31, 2011
Inspectors: D. Dumbacher, Senior Resident Inspector (Team Lead)
B. Correll, Reactor Inspector
J. Dixon, Senior Resident Inspector
J. Drake, Senior Reactor Inspector
N. Makris, Project Engineer
L. Willoughby, Senior Project Engineer
Approved By: G. Miller, Branch Chief
Division of Reactor Projects
1 Enclosure
SUMMARY OF FINDINGS
IR 05000482/2011006, 02/07- 03/31/2011, Wolf Creek Generating Station, Supplemental
Inspection - Inspection Procedure 95002.
This supplemental inspection was conducted by two senior resident inspectors, a reactor
inspector, a senior reactor inspector, a project engineer, and a senior project engineer. One
Green noncited violation was identified. The significance of most findings is indicated by their
color (Green, White, Yellow, or Red) using Inspection Manual Chapter 0609, "Significance
Determination Process." The crosscutting aspect is determined using Inspection Manual
Chapter 0310, "Components Within the Cross Cutting Areas." Findings for which the
significance determination process does not apply may be Green or be assigned a severity level
after NRC management review. The NRC's program for overseeing the safe operation of
commercial nuclear power reactors is described in NUREG-1649, "Reactor Oversight Process,"
Revision 4, dated December 2006.
Cornerstones: Initiating Events and Mitigating Systems
The NRC staff performed this supplemental inspection in accordance with Inspection
Procedure 95002, "Supplemental Inspection for One Degraded Cornerstone or Any Three White
Inputs in a Strategic Performance Area," to assess the licensee's evaluations associated with
White performance indicators for unplanned scrams per 7000 critical hours, safety system
functional failures, and unplanned scrams with complications. Inspection Procedure 92723,
"Follow Up Inspection for Three or More Severity Level IV Traditional Enforcement Violations in
the Same Area in a 12 Month Period," was also performed.
The inspectors determined that the Wolf Creek staff performed a comprehensive evaluation of
the events that led to the degraded Initiating Events Cornerstone and three white inputs in the
reactor safety strategic performance area. Wolf Creek's evaluation identified root causes of the
collective issues to be related to: 1) inadequate management oversight/standards enforcement,
2) lack of knowledge concerning the components of nuclear safety culture and crosscutting
issues, and 3) inadequate hardware monitoring.
In addition to assessing the licensee's evaluations, the inspection team performed an
independent extent-of-condition and extent-of-cause review and a focused inspection of the site
safety culture as it related to the root cause evaluations. The team concluded that the Wolf
Creek root cause evaluations and corrective actions, both completed and planned, addressed
the extent-of-condition and extent-of-cause, determined if safety culture contributed to the issue,
and established and scheduled corrective actions that are sufficient to address the causes and
prevent recurrence of the White performance indicators.
Based on independent inspection, the team also determined that the licensee's assessment of
Wolf Creek's safety culture was accurate and reflected the conditions at the site.
The root cause evaluations appropriately identified needed improvements associated with
safety culture behaviors.
2 Enclosure
A. NRC-Identified or Self-Revealing Findings
Cornerstone: Mitigating Systems
Green. The inspectors identified a noncited violation of 10 CFR Part 50, Appendix B,
Criterion V, "Instructions, Procedures, and Drawings," for the failure to follow Procedure
AP 28A-0100, "Condition Reports," Revision 13. On February 17, 2011, the licensee
received laboratory test results on the emergency diesel generator B fuel oil storage tank
and determined that the cloud point parameter was out of specification at -8° Celsius.
However, Procedure AP 28A-0100, step 5.13.3, required the licensee to evaluate
condition report data to identify and evaluate potential trends. The emergency diesel fuel
oil storage tank cloud point parameter had been trending closer to the acceptance
criteria over the last several fuel oil additions. The licensee had allowed the original fuel
oil vendor to continue to deliver fuel that was out of specification which resulted in a
gradual trend toward the limits of the chemistry parameters. This trend was not
appropriately evaluated because the licensee had not performed training to ensure that
consistent and appropriate evaluations would be performed.
This finding was more than minor because it affected the Mitigating Systems
Cornerstone attribute of equipment performance by impacting the cornerstone objective
to ensure the availability, reliability, and capability of systems that respond to initiating
events to prevent undesirable consequences. This deficiency directly resulted in
emergency diesel generator B being declared inoperable due to its fuel oil storage tank
being out of specification. The inspectors performed the significance determination
using NRC Inspection Manual Chapter 0609.04, "Phase 1 - Initial Screening and
Characterization of Findings," because it affected the Mitigating Systems Cornerstone
while the plant was at power. The finding was determined to be 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; it did not represent the loss of a single train
for greater than technical specification allowed time; it did not represent a loss of one or
more non-technical specification risk-significant equipment for greater than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />;
and it did not screen as potentially risk significant due to seismic, flooding, or severe
weather. In addition, this finding had a human performance crosscutting aspect
associated with resources in that the licensee did not ensure that the corrective action
program coordinators were effectively trained to cognitively and analytically trend
condition reports H.2(b)(Section 4OA4).
B. Licensee-Identified Violations
A violation of very low safety significance, which was identified by the licensee, has been
reviewed by the inspectors. Corrective actions taken or planned by the licensee have
been entered into the licensee's corrective action program. This violation and
associated condition report numbers are listed in Section 4OA7.
3 Enclosure
REPORT DETAILS
4. OTHER ACTIVITIES
4OA3 Event Follow-up
.1 (Closed) Licensee Event Report 05000482/2010-005-00: Reactor Trip due to Low
Steam Generator Level from Trip of Main Feedwater Pump
On March 2, 2010, a trip of the Train A main feedwater pump caused a low-low steam
generator water level reactor trip. The trip of the main feedwater pump was due to
nonsafety-related inverter PN09 failing to transfer to the alternate supply during
preparation for minor maintenance. During performance of Procedure SYS PN-200,
"Energizing and Deenergizing Inverters PN09 and PN10," inverter PN09 failed to
transfer from the normal to alternate power supply due to sticking of the reed relay on
the static transfer switch circuit board after exceeding its design life. The failure to
transfer caused a loss of speed signal to main feedwater pump A resulting in an
overspeed trip and caused a loss of steam dump capability. The unit received a
feedwater isolation signal and an auxiliary feedwater actuation signal.
The inspectors determined the licensee's root cause evaluation inappropriately identified
the direct cause as the root cause and incorrectly stated the actions taken to replace the
cards were sufficient such that no corrective actions to prevent recurrence were
necessary. The NRC inspectors concluded that a contributing cause identified in the
root cause evaluation, the decision to continue operating with equipment beyond its
design life, was more appropriate as a root cause. The inspectors also identified that the
root cause evaluation for extent-of-condition was narrowly focused, in that the licensee
only identified other inverters as being within the extent-of-condition. A broader extent-
of-condition would have included any electronic circuit boards with design life limitations
and would not be limited to inverters. Corrective actions taken by the licensee to
address the contributing cause included replacing the circuit cards and preventative
maintenance frequency changes to prevent exceeding the design life of the circuit cards.
The inspectors determined the corrective actions were appropriate to prevent
recurrence.
The inspectors concluded the failure to identify the lowest level root cause and the
narrowly focused extent-of-condition determination were a violation of 10 CFR Part 50,
Appendix B, Criterion V, "Instructions, Procedures, and Drawings," involving a failure to
follow root cause Procedure AI 28A-001, "Level 1 CR Evaluation (IIT)," Revision 12.
Since the licensee took appropriate corrective actions as part of an identified contributing
cause, the inspectors determined this violation was of minor safety significance.
Violations of minor safety or security concern generally do not warrant enforcement
action but must be corrected.
Event follow-up inspections by NRC inspectors identified two Green findings associated
with this event: FIN 05000482/2010002-04 and NCV 05000482/2010002-05, The
inspectors reviewed the licensee event report and determined that the report adequately
documented the summary of the event, including the potential safety consequences,
cause of the event, and corrective actions required to address the performance
deficiency. No additional findings were identified. This licensee event report is closed.
4 Enclosure
.2 (Closed) Licensee Event Report 05000482/2010-012-00: Reactor Trip due to Operator
Inability to Control Steam Generator Level Oscillations at Low Power
On October 17, 2010, a reactor trip occurred due to inadequate steam generator water
level control during low power operations. During plant startup following a forced
outage, reactor power was increased above 10 percent to approximately 17 percent
when starting to roll the main turbine and synchronize to the grid. While rolling the main
turbine, feedwater temperature began to drop due to insufficient feedwater preheating.
The operators took manual control of steam generator water level but were unable to
maintain level below the high-high steam generator water level turbine trip and
feedwater isolation signal. Auxiliary feedwater was unable to match steam demand and
the reactor tripped on low-low steam generator water level. The inability to control
steam generator water level was due to the decreased feedwater temperature caused by
insufficient feedwater preheating during the power increase associated with placing the
main turbine on line. Operators failed to recognize that feedwater preheating from the
main steam system had a capacity limit of 10 percent until the main turbine is brought on
line.
The licensee submitted a licensee event report for the reactor trip on December 16,
2010. The licensee made procedure changes to require the main turbine to be
synchronized to the grid prior to exceeding 10 percent reactor power. Event follow-up
inspections by NRC inspectors identified four Green noncited violations associated with
this event;05000482/2010005-08; 05000482/2010005-09; 05000482/2010005-10; and
05000482/2010005-11. The inspectors reviewed the licensee event report and
determined that the report adequately documented the summary of the event including
the potential safety consequences, cause of the event, and corrective actions required to
address the performance deficiency. No additional findings were identified. This
licensee event report is closed.
4OA4 Supplemental Inspection (95002)
.01 Inspection Scope
The NRC staff performed this supplemental inspection in accordance with Inspection
Procedure 95002, "Inspection for One Degraded Cornerstone or Any Three White Inputs
in a Strategic Performance Area," to assess the degraded Initiating Events Cornerstone
and the three White inputs to the reactor safety strategic performance area. The team
also performed Inspection Procedure 92723, "Follow Up Inspection for Three or More
Severity Level IV Traditional Enforcement Violations in the Same Area in a 12 Month
Period," in conjunction with the supplemental inspection. The inspection objectives were
to:
- provide assurance that the root and contributing causes of risk-significant issues
were understood
- provide assurance that the extent-of-condition and extent-of-cause of risk-
significant issues were identified and to independently assess the extent-of-
condition and extent-of-cause of individual and collective risk-significant issues
- independently determine if safety culture components caused or significantly
contributed to the risk significant issues
5 Enclosure
- provide assurance that the licensee's corrective actions for risk-significant issues
were or will be sufficient to address the root and contributing causes and to
preclude repetition
The licensee entered the Degraded Cornerstone Column of the NRCs Action Matrix in
the first quarter of 2010 as a result of three performance indicators crossing the
threshold from Green (very low safety significance) to White (low to moderate safety
significance). The performance indicators were Unplanned Scrams per 7000 Critical
Hours, Unplanned Scrams with Complications, and Safety System Functional Failures.
The licensee staff informed the NRC that Wolf Creek was prepared for the supplemental
inspection on October 29, 2010. To determine the causes and organizational attributes
that resulted in the three White performance indicators, the licensee performed root
cause evaluations documented in Condition Reports 26805 (collective), 23119 (safety
system functional failures), 24445 (unplanned scrams) and 25817 (scrams with
complications). These condition reports were associated with many individual event
condition reports. Altogether the inspection scope and the licensee actions included well
over 1000 corrective actions. The team noted that the licensee recovery team
performed an overarching safety culture review to determine whether safety culture
components and aspects contributed to the performance issues that led to the White
NRC performance indicators. The team inspected this effort by reviewing Condition
Reports 23032 and 25896. For the traditional enforcement violations, the team reviewed
licensee efforts documented in Condition Report 23110. The inspection team reviewed
the licensee's root cause and other supporting evaluations, and the team reviewed
corrective actions that were taken or planned to address the identified causes. The
inspection team also held discussions with licensee personnel to ensure that the root
and contributing causes and the contribution of safety culture components were
understood and corrective actions taken or planned were appropriate to address the
causes and preclude repetition. The inspection team independently assessed the
extent-of-condition and extent-of-cause of the identified issues and performed an
assessment of whether any safety culture components caused or significantly
contributed to the issues.
.02 Evaluation of the Inspection Requirements
02.01 Problem Identification
a. Identification of the issue (i.e. licensee-identified, self-revealing, or NRC-identified) and
the conditions under which the issue was identified
As a result of the multiple reactor scrams and safety system functional failures in 2009
and 2010, the licensee identified three White performance indicators through the NRC's
performance indicator reporting process.
The Unplanned Scrams with Complications performance indicator crossed the threshold
from Green to White as a result of unplanned scrams in 2009 that were the subject of a
Reactor Oversight Process Working Group Frequently Asked Question. On April 28,
2009, the main feedwater regulating valve controller power supply fuses failed, isolating
flow to steam Generator B and resulting in a reactor trip from loss of power to a main
feed regulating valve controller. Also, on August 19, 2009, a complete loss of offsite
power resulted in a complicated scram. Based on the resolution of the Frequently Asked
6 Enclosure
Question, Wolf Creek reported both of these reactor trips as Unplanned Scrams with
Complications, which caused this performance indicator in the Initiating Events
cornerstone to be White starting in the third quarter of 2009.
On March 8, 2010, during a plant start up at approximately 42 percent power, operators
manually tripped the reactor following an unplanned trip of the only running feedwater
pump. This plant startup was being conducted following a previous reactor trip on
March 2, 2010, in which loss of power to an electrical inverter led to a trip of main
feedwater Pump A and resultant low steam generator water levels. These reactor trips
combined with two others from April 2009 and August 2009 caused the Unplanned
Scrams per 7000 Critical Hours performance indicator to be White. Additionally, in
April 2010 Wolf Creek reported four safety system functional failures for the first
calendar quarter of 2010. Combined with the five others previously reported, these
functional failures caused the Safety System Functional Failures performance indicator
also to be White.
The inspectors verified that this information was appropriately documented in the
licensees evaluations.
b. Issue duration and prior opportunities for identification
The degraded Initiating Events Cornerstone and the three White inputs to the reactor
safety strategic performance area existed from March 2010 when they were identified by
Wolf Creek's performance indicator submittals. The Complicated Scrams performance
indicator crossed the Green/White threshold in the third quarter 2009. The Unplanned
Scrams per 7000 Critical Hours performance indicator was White starting in the first
quarter of 2010. Both of these performance indicators returned to Green in the second
quarter of 2010. The Safety System Functional Failures performance indicator also first
crossed the Green/White threshold in the first quarter of 2010 and did not return Green
until after the first quarter of 2011. Each of the reactor scrams and safety system
functional failures was an opportunity to identify the need for corrective actions to
reverse the negative performance trend.
The inspectors concluded that the licensees evaluations adequately identified how long
each issue existed and prior opportunities for identification of the failures.
c. Licensee documentation of the plant specific risk consequences, as applicable, and
compliance concerns associated with the issues both individually and collectively
The inspectors verified that the licensee's evaluation adequately documented the plant
specific risk consequences in qualitative statements that equipment failures directly
affect nuclear safety by challenging critical safety functions and operator response.
There were no previously documented findings associated with the scrams or safety
system functional failures that were more than very low safety significance.
02.02 Root Cause, Extent-of-Condition, and Extent-of-Cause Evaluation
a. Determine that the licensee evaluated the issue using a systematic methodology to
identify the root and contributing causes
The licensee used the following methods to complete the root cause evaluation:
7 Enclosure
- event and causal factor charting
- hazard-barrier-target analysis
- management oversight and risk tree (MORT) analysis
- fault tree analysis
The NRC team concluded the licensee evaluated the issues using systematic
methodologies to identify root and contributing causes.
b. Determine that the licensee's root cause evaluation was conducted to a level of detail
commensurate with the significance of the issue
The licensees evaluation identified the root causes of collective issues to be:
1) inadequate management oversight, 2) lack of knowledge concerning components of
nuclear safety culture, and 3) inadequate hardware monitoring. The NRC team
performed a focused inspection to independently assess the validity of the licensee's
conclusions regarding the extent-of-condition and extent-of-cause of the issues. The
NRC inspection team review for each performance indicator, individually and collectively,
determined that the licensees root cause evaluation level of detail was commensurate
with the significance of the problem.
c. Consideration of prior occurrences of the issue and knowledge of operating experience
Based on the licensee's detailed evaluation and conclusions, the inspection team
determined that the licensee's root cause analysis included an appropriate consideration
of prior occurrences of the issue and knowledge of prior operating experience.
d. Determine that the licensee's root cause evaluation addresses the extent-of-condition
and extent-of-cause of the issues
The inspectors concluded that the licensees root cause analysis appropriately
addressed the extent-of-condition and the extent-of-cause of the issue. However, for
many of the root cause evaluations the documentation was high-level, difficult to follow,
and did not always provide a strong basis for implementation and closure of the
individual corrective actions. The team determined that, in these cases, appropriate
corrective actions were specified for each root and contributing cause in other condition
reports.
e. Review the licensee's root cause, extent-of-condition, and extent-of-cause evaluations in
order to verify that the licensee appropriately considered the safety culture components
as described in Inspection Manual Chapter 0305
Because multiple condition reports and several safety culture aspects were associated
with the performance issues, the licensee conducted collective reviews of the past two
safety culture assessments and the six significant contributing condition reports. This
effort resulted in the licensee creating roll-up Condition Report 26805 to prioritize safety
culture corrective actions in September 2010. The team concluded that the prioritization
was logical and that the corrective actions, while appropriate, needed increased
oversight and reinforcement. In response to the inspection teams observations, the
licensee added specific items to improve safety culture behaviors, trending, and
knowledge levels of the operating and engineering departments to the Recovery Change
8 Enclosure
Management Plan for the upcoming cycle. The licensee also strengthened the safety
culture communication plan to reinforce human performance tool usage at all levels.
02.03 Corrective Action
a. Determine that: 1) the licensee specified appropriate corrective actions for each root
and/or contributing cause, or 2) an evaluation that states no actions are necessary is
adequate
The licensee initiated well over 1000 corrective actions to address the root and
contributing causes from the individual and collective cause evaluations. The inspectors
concluded that the licensee had specified appropriate corrective actions for each root
and/or contributing cause. The inspectors observed some cases where the licensees
corrective actions could have been more specifically tied to causes and more generally
applied to safety culture aspects. These observations are discussed in Section 02.07 of
this report.
b. Determine that the licensee prioritized corrective actions with consideration of risk
significance and regulatory compliance
The majority of the corrective actions developed by the licensee involved long term
maintenance plans and plant modifications. The licensee also prioritized communication
of standards and newly formed additional review processes as short term items. These
short term items were still in progress at the time of the inspection. The inspection team
determined that there were no risk significant immediate corrective actions necessary.
The inspection team concluded that the corrective actions identified in the root cause
evaluations for the White performance indicators were appropriately prioritized based on
risk significance and regulatory compliance.
c. Determine that the licensee established a schedule for implementing and completing the
corrective actions
The inspection team found that the licensee's root cause evaluations established many
different, independent schedules for completion of the over 1000 corrective actions.
Tracking, evaluating and closing corrective actions was assigned to the licensee
recovery team. The NRC inspection team observed that the individual schedules did not
appear to be coordinated with one another. To address the inspection teams
observation, the licensee staff compiled and provided a table to the inspectors which
tracked each corrective action item milestone with its corresponding completion date.
The inspectors concluded the revised schedule was appropriate for effectively
implementing and completing the corrective actions.
d. Determine that the licensee developed quantitative and/or qualitative measures of
success for determining the effectiveness of the corrective actions to preclude repetition
The measures developed by the licensee for determining the effectiveness of corrective
actions included the following:
- Corrective Action and Operating Experience Review Board external reviews
- Increased frequency of quality assurance audits to assess the adequacy of the
corrective action program initiatives generated
9 Enclosure
- Increased frequency of safety culture assessments
The inspection team determined that the quantitative and qualitative measures
developed by the licensee for determining the effectiveness of the corrective actions
were appropriate.
e. Determine that the licensee's planned or taken corrective actions adequately address a
Notice of Violation that was the basis for the supplemental inspection, if applicable
A Notice of Violation was not the basis for this supplemental inspection.
02.04 Independent Assessment of Extent-of-Condition and Extent-of-Cause
a. Inspection Scope
Inspection Procedure 95002 requires that the inspection staff perform a focused
inspection to independently assess the validity of the licensee's conclusions regarding
the extent-of-condition and extent-of-cause of the issue. The objective of this
requirement is to independently sample performance, as necessary, within the key
attributes of the cornerstones that are related to the subject issue to ensure that the
licensee's evaluation regarding the extent-of-condition and extent-of-cause is sufficiently
comprehensive.
The inspectors conducted independent extent-of-condition and extent-of-cause reviews
for the issues associated with the White performance indicators. The inspection staffs
independent review focused on the primary root causes associated with the performance
indicators in addition to the licensees identified contributing causes that involved more
specific aspects of the broader root causes. The inspection staff assessed whether the
licensees extent-of-condition and extent-of-cause evaluations sufficiently identified and
bounded all engineering and maintenance organizational issues. The staff also
assessed whether the licensees extent-of-condition and extent-of-cause evaluations
sufficiently determined the actual extent of similar organizational issues that potentially
existed in other station departments, programs, and processes. The team independently
sampled performance within the key attributes of the Initiating Events and Mitigating
Systems Cornerstones that are related to the contributors of the performance issues to
ensure that the licensee's evaluation regarding the extent-of-condition and extent-of-
cause were sufficiently comprehensive.
In conducting this independent review, the inspection staff interviewed station
management and personnel, reviewed program and process documentation, and
reviewed existing station program monitoring and improvement efforts, including review
of corrective action documents.
b. Assessment
The team concluded that the licensee had identified all substantive extent-of-condition
and extent-of-cause issues. However, the teams independent extent-of-condition and
extent-of-cause review identified some cases where the licensees evaluations were
narrowly focused. For example, the licensee's evaluation of scrams with complications
in Condition Report 25817 limited the review of main feedwater system health to just the
startup feedwater pump, which inappropriately excluded many components with the
10 Enclosure
potential to affect system performance. The evaluation also did not evaluate the
August 2009 loss of offsite power and condensate and heater drain bus event, making
instead a statement that a loss of offsite power will always result in a complicated scram.
The team identified this as another example where the evaluation of Condition
Report 25817 missed an opportunity to improve feedwater system reliability. The team
concluded Wolf Creek's root cause analysis procedures could be improved to enable the
licensee to consistently identify systemic causal factors.
As a result of the inspection teams observations, Wolf Creek reviewed Condition
Report 25817 to identify additional interim and long-term corrective actions. This
included a review by the Quality group and bringing in additional root cause evaluators
to ensure the root cause analysis procedures were improved as needed.
02.05 Safety Culture Consideration
a. Inspection Scope
Inspection Procedure 95002 requires that the inspection team perform a focused
inspection to independently determine that the licensee's root cause evaluation
appropriately considered whether any safety culture component caused or significantly
contributed to any risk significant issue.
The inspection team reviewed condition reports and procedures and conducted
interviews with licensee personnel to determine if the licensee properly considered
whether any safety culture component caused or contributed to the performance issues.
Additionally, the inspectors performed a review of the common cause evaluation.
b. Assessment
As part of the collective root cause evaluation, the licensee evaluated the identified root
and contributing causes against the safety culture components that could have
contributed to the issues. The licensee's root cause evaluation included a discussion of
the 13 safety culture components as described in Regulatory Issue Summary 2006-013,
"Information on the Changes Made to the Reactor Oversight Process to More Fully
Address Safety Culture."
The inspection team independently confirmed the licensees conclusion that improving
safety culture behaviors should be a high priority item for the recovery effort. The
documented station reviews indicated that every safety culture component was a
contributor to the performance issues, and all were significant contributors with the
exception of self- and independent-assessments. The inspection team concluded all the
safety culture components were significant contributors. The inspection team confirmed
that the licensee established appropriate corrective actions to address safety culture.
The team identified challenges to Wolf Creek's ensuring long-term promotion of a
positive safety culture. Specifically:
- Although safety concept is a recognized value in the organization, it is
inconsistently accepted and understood across all levels of personnel. Some
problems still exist in the transmission, comprehension, and implementation of
the safety message.
11 Enclosure
- Some individuals readily accept responsibility for and take ownership of
problems, while others are still reluctant to do so.
- Observed safety behaviors were not consistently integrated into all activities in
the organization. Processes and programs are in various stages of transition,
which often reduces their effectiveness.
- An integrated and cohesive organizational safety leadership process does not yet
exist. The values and attitudes of the workforce are generally positive, but the
team identified that personnel are not yet aligned with a common set of values.
02.06 Evaluation of Inspection Manual Chapter 0305 Criteria for Treatment of Old Design
Issues
The licensee did not request credit for self-identification of an old design issue; therefore,
the risk-significant issue was not evaluated against the Inspection Manual Chapter 0305
criteria for treatment of an old design issue.
02.07 Findings and Observations
a. Introduction. The inspectors identified a Green noncited violation of 10 CFR Part 50,
Appendix B, Criterion V, "Instructions, Procedures, and Drawings," for the failure to
follow Procedure AP 28A-0100, "Condition Reports," Revision 13.
Description. On February 17, 2011, the licensee received laboratory test results on the
Train B emergency diesel generator fuel oil storage tank and determined that the cloud
point parameter was out of specification at -8° Celsius. The specification limit for cloud
point was no higher than -9° Celsius. The licensee subsequently declared the
emergency diesel generator inoperable and entered Technical Specification 3.8.3. As
part of the review of the event, the licensee sent an additional sample from the fuel oil
storage tank to the same laboratory, as well as to an additional laboratory for
comparison. The licensee also sent samples from the Train A emergency diesel
generator fuel oil storage tank to determine the extent-of-condition. These actions are
documented in Condition Report 33750.
Procedure AP 28A-0100, step 5.13.3, required licensee personnel to evaluate condition
report data to identify and evaluate potential trends. The emergency diesel fuel oil
storage tank cloud point parameter had been trending closer to the acceptance criteria
over the last several fuel oil additions. In various condition reports over the past two
years, the licensee documented that the cloud point parameter had been out of
specification in new fuel oil shipments. In addition to the sample in February 2011,
Condition Reports 21044, 25018 and 26345 documented the cloud point parameter
being an issue in October 2009, April 2010 and June 2010, respectively. Condition
Report 26345 did not receive an appropriate review to identify that the adverse trend, if
not resolved promptly, could result in the emergency diesel generator becoming
inoperable. Corrective actions from Condition Report 26345 included purchasing fuel oil
from a new vendor that would provide a low cloud point, but this was not implemented in
a timely manner to prevent the unplanned technical specification entry. As a result, on
February 17, 2011, the Train B fuel oil storage tank cloud point parameter went out of
specification requiring the licensee to withdraw fuel oil and replace it with in-specification
fuel oil from the new vendor. Two tanker loads of new fuel oil were placed into the
12 Enclosure
storage tank and a multilevel sample of the resulting mixture was analyzed to ensure
that all chemistry parameters were within specification. The licensee had allowed the
original fuel oil vendor to continue to deliver fuel that was out of specification which
resulted in a gradual trend toward the limits of the chemistry parameters. This trend was
not appropriately evaluated because the licensee had not performed training to ensure
that consistent and appropriate evaluations would be performed.
Analysis. Failure to track and trend the emergency diesel generator chemistry
parameters as required by the corrective action program procedure was a performance
deficiency. The finding was more than minor because it affected the Mitigating Systems
Cornerstone attribute of equipment performance by impacting the cornerstone objective
to ensure the availability, reliability, and capability of systems that respond to initiating
events to prevent undesirable consequences. This deficiency directly resulted in
emergency diesel generator B being declared inoperable due to its fuel oil storage tank
being out of specification. The inspectors performed the significance determination
using NRC Inspection Manual Chapter 0609.04, "Phase 1 - Initial Screening and
Characterization of Findings," because it affected the Mitigating Systems Cornerstone
while the plant was at power. The finding was determined to be 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; it did not represent the loss of a single train
for greater than technical specification allowed time; it did not represent a loss of one or
more non-technical specification risk-significant equipment for greater than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />;
and it did not screen as potentially risk significant due to seismic, flooding, or severe
weather. In addition, this finding had human performance crosscutting aspects
associated with resources in that the licensee did not ensure that the corrective action
program coordinators were effectively trained to cognitively and analytically trend
condition reports H.2(b).
Enforcement. Title 10 CFR Part 50, Appendix B, Criterion V, "Instructions, Procedures,
and Drawings," requires, in part, that activities affecting quality shall be prescribed by
documented procedures of a type appropriate to the circumstances and shall be
accomplished in accordance with these procedures. Procedure AP 28A-0100,
"Condition Reports," step 5.13.3 requires the licensee, in part, to evaluate condition
report data to identify and evaluate potential trends. Contrary to this, from October 2009
to February 19, 2011, the licensee failed to evaluate condition report data to identify and
evaluate potential trends in emergency diesel generator fuel oil storage tank chemistry
parameters. As a result, the station entered Technical Specification 3.8.3 for a high
cloud point on the Train B emergency diesel generator fuel oil storage tank. Immediate
corrective actions included withdrawing fuel oil and replacing it with new fuel oil until the
cloud point could be reduced to below the maximum value. Since this violation was of
very low safety significance and was documented in the licensee's corrective action
program as Condition Reports 33395, 33435, and 33750, it is being treated as a
noncited violation consistent with Section 2.3.2 of the NRC Enforcement Policy:
NCV 05000482/2011006-01, "Failure to Trend Emergency Diesel Generator Chemistry
Parameters Results in an Unplanned Technical Specification Entry."
13 Enclosure
b. NRC Team Observations
1) Maintenance to Improve Equipment Reliability
The inspection team noted that the licensees corrective actions for improving equipment
reliability focused primarily on preventative maintenance improvements. The team
determined this focus initially may not result in protecting the functionality of the key
systems inputting to the safety system functional failure performance indicator in the
short term. The team concluded that a broader root cause of inadequate maintenance in
general, versus just improved preventative maintenance, may be more appropriate to
address equipment reliability issues. This action is consistent with the licensee roll-up
root cause of tolerance for known degraded equipment conditions. The team determined
that the recent refueling water storage tank and emergency diesel generator fuel oil
storage tank degraded chemistry issues, emergency diesel generator fuel rack pin
inoperability, and component cooling water system voiding were examples that
demonstrated the importance of improvement in general maintenance practices and
represented conditions that could challenge the safety system functional failure
performance indicator.
The team concluded that corrective actions to develop a preventative maintenance
optimization plan, improve operating experience reviews, and perform significant main
feedwater modifications were appropriate. However, the corrective actions to add digital
feedwater controls will not be complete until Fall 2015.
The licensee identified inadequate equipment performance monitoring and trending as a
maintenance improvement item, but initiated actions for "new" systems/components
only. The team concluded a broader application of this action would be more
appropriate. For example, the team identified that the station thermography tool is not
being used to its full extent. This was similar to the previous limited use of ultrasonic
testing and guided wave technology for essential service water corrosion issues. The
licensee initiated Condition Report 33435 to evaluate additional corrective actions for
equipment performance and monitoring.
2) Corrective Action Documentation
The team identified several instances where the documentation of corrective actions was
not clearly defined through the corrective action program, which could provide
challenges to the timely completion of the actions and to the ability of the licensees
quality control organization to perform effectiveness reviews. For example, corrective
actions for root causes in Condition Report 23119 (safety system functional failures)
were contained in Condition Report 24445, but were not listed as corrective actions to
prevent recurrence in Condition Report 24445. This could lead to closure of the actions
out without sufficient reviews to ensure effective corrective actions occurred. As a
second example, overall station roll-up Condition Report 26805 contained no
documented corrective actions. Instead, this condition report provided a prioritization of
the common causes in the other six high level individual event roll-up condition reports
(see Attachment 2). The documentation of the corrective actions for the causes and
safety culture component concerns are embedded in various other condition reports.
The team identified this as a challenge to correctly implementing and closing out the
individual corrective actions. The licensee took actions to address this issue, including
initiating Condition Reports 33722 and 33958 and developing additional root cause
14 Enclosure
evaluators to improve documentation standards and reduce root cause evaluation
backlogs.
The team also identified that two key licensee initiatives (each a corrective action to
prevent recurrence), the Preventative Maintenance Optimization Plan and the Ops
Focus Plan, were essentially mission statements and lacked the details necessary to
guide implementation. As a result of the team inspection, the licensee developed
revisions to these plans to provide better detail and clarity.
3) Training as a Corrective Action
The licensee identified a corrective action of improved training, modeled after previous
changes made to improve the technical program for engineers and operators, to address
performance issues associated with both the safety system functional failures and
unplanned scrams performance indicators. However, training provided as a corrective
action for risk assessments failed the licensees initial effectiveness review. This was
one of the first effectiveness reviews performed by the licensee, and it identified that the
training corrective action was narrowly focused and not likely to reach everyone affected.
The team identified some additional examples where training as a corrective action had
not yet been fully effective, including:
- A corrective action review board graded the root cause evaluation of "inadequate
clearances orders due to not isolating the energy source," as acceptable without
a training corrective action when the cause was an inability to read prints.
- At the time of the inspection, neither engineering nor training departments had
been trained on the updated operating experience process.
- The team observed some cases where system engineer knowledge levels were
not broad or integrated.
- Root cause team member training was not consistently producing thorough
extent-of-cause results.
- The team received interview comments indicative of operations and training
department dissatisfaction with support by the other department. Other
comments also identified cases where managers may have decided not to assign
training related corrective actions based on limited training resources.
The team concluded that broader training, improvements for both engineers and
operators is needed. The licensee initiated Condition Reports 34280 and 34281 to
address engineering and operations department training issues.
4) Problem Identification
The team concluded the various review board initiatives, specifically the corrective action
review board challenge meetings, apparent cause evaluation level corrective action and
collegial operating experience review boards, were positive efforts. However, individual
level behavior changes will be needed to identify trends or deficiencies in equipment and
engineering performance. The team noted some cases where the threshold for
15 Enclosure
identifying and trending degraded conditions in the plant could be improved. Examples
include:
- A high level alarm indicative of actual rising essential service water vault water
level was initially assumed to be invalid in Fall 2010. This resulted in delaying an
operability assessment over two shifts.
- The team identified that self- and independent-assessments were not being
effectively used to identify and correct problems. The licensee initiated condition
report 34076 and directed the quality assurance department to perform an
additional surveillance of the corrective action program to address this issue.
- The team noted two examples of initial failure to act on contracted engineering
evaluations. The first example involved a contractor evaluation of the feedwater
pump suction strainer in December 2010 that stated the strainer could introduce
new failure mechanisms. The inspection team noted a failed strainer could
release debris which could impact the feedwater regulating and isolation valves.
The licensee initiated Condition Report 32445 to evaluate this condition. The
second example was previously identified in an NRC inspection of essential
service water and involved a deficient in-house engineering analysis accepted
without action despite outside contractor evaluations stating that water hammer
stresses were significant enough to warrant inclusion in the system design
calculations. At the close of the inspection the licensee was reevaluating both of
these issues.
- The team identified some process programs that may result in tracking and
correcting problems outside of the corrective action program. One example is
the PILOT system used to record and trend management field observations. The
licensee initiated Condition Report 33316 to address this concern.
5) Management Oversight and Leadership
In Condition Report 26805, "Collective Significance of Degraded Cornerstone
Performance," the licensee identified Management/Oversight/Standards Enforcement as
one of the overall root causes of the sites performance issues. The team concluded
that licensee actions to address improving management oversight were appropriate.
However, the team identified that the additional review boards requiring management
participation had the unintended consequence of reducing the amount of time available
for managers and supervisors to conduct plant tours and field observations. The team
also identified some cases where review boards accepted quality assurance reports of
effective program performance despite identified repeat findings, and cases where the
review boards did not consistently challenge extent-of-condition and extent-of-cause
issues in root cause evaluations.
To address the teams observations the licensee developed initiatives to improve
standards for leadership meetings, division manager alignment meetings and plant wide
communication efforts.
16 Enclosure
4OA6 Meetings
Exit Meeting Summary
On April 5, 2011, the inspectors presented the inspection results to Mr. M. Sunseri and
other members of the licensee staff. The licensee acknowledged the issues presented.
The inspectors identified that proprietary information was reviewed but would not be
retained following report issuance or included in the inspection report.
4OA7 Licensee-identified Violations
The following violation of very low safety significance (Green) was identified by the
licensee and is a violation of NRC requirements which meets the criteria of Section 2.3.2
of the NRC Enforcement Policy, NUREG-1600, for being dispositioned as a noncited
violation.
- Title 10 CFR 50.65 a(4) requires, in part, that before performing maintenance
activities, the licensee shall assess and manage the increase in risk that may
result from the proposed maintenance activities. Contrary to the above, during
the weeks of November 29, 2010, December 27, 2010 and January 17, 2011,
Wolf Creek failed to properly identify and take appropriate risk management
actions for medium and high risk maintenance activities as required by station
Procedure AP 22C-007, "Risk Management and Contingency Planning,"
Revision 4. The inspectors performed the significance determination using NRC
Inspection Manual Chapter 0609.04, "Phase 1 - Initial Screening and
Characterization of Findings," and Manual Chapter 0609, Appendix K,
"Maintenance Risk Assessment and Risk Management Significance
Determination Process" and determined the finding was of very low safety
significance (Green) because it related only to risk management actions and did
not result in an increase in core damage probability. This licensee entered this
issue into the corrective action program as Condition Reports 00032886
and 00032887.
17 Enclosure
SUPPLEMENTAL INFORMATION
KEY POINTS OF CONTACT
Licensee Personnel
L. Bell, Systems Engineering
M. Blow, Operations
S. Hedges, Site Vice President
D. Hooper, Supervisor, Licensing
T. Jensen, Manager, Chemistry
S. Koenig, Manager Corrective Action
W. Norton, Manager IPS/Scheduling
L. Parmenter, Assistant to Manager, Operations Department
G. Pendergrass, Director, Plant Engineering
L. Ratzlaff, Supervisor, Support Engineering
E. Ray, Manager Quality
L. Rockers, Licensing Engineer
R. Smith, Plant Manager
M. Sunseri, President and Chief Executive Officer
S. Wahlmeier, Systems Engineering
J. Yunk, Manager, Human Resources
LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED
Opened and Closed
05000482/2011006-01 NCV Failure to Trend Emergency Diesel Generator Chemistry
Parameters Results in an Unplanned Technical Specification
Entry (Section 4OA4)
Closed
05000482/2010-005-00 LER Reactor Trip due to Low Steam Generator Level from Trip of
Main Feedwater Pump (Section 4OA3)
05000482/2010-012-00 LER Reactor Trip due to Operator's Inability to Control Steam
Generator Level Oscillations at Low Power (Section 4OA3)
LIST OF DOCUMENTS REVIEWED
Section 4OA4: Supplemental Inspection
PROCEDURES
NUMBER TITLE REVISION
AI 09A-008 Engineering Allocation 3
AI 16C-007 Work Order Planning 30
A1-1 Attachment 1
AI 17C-005 Shift Manager Selection, Initial Training and Continuing 5
Training Program
AI 18A-001 Receipt, Investigation, and Closure of Employee Concerns 2
AI 18A-002 Conducting Exit/Walk-in Interviews 1A
AI 18A-003 Preparation, Maintenance, and Security of Employee 1A
Concerns Files
AI 21-016 Operator Time Critical Actions Validation 1
AI 22C-008 Work Scoping Team 6
AI 22C-012 Quality Review Team (QRT) for Maintenance Work Planning 0
AI 22D-001 High Impact Teams 5
AI 22I-004 Project Ranking Guide 0, 1, 2
AI 23M-003 Maintenance Rule Expert Panel Duties and Responsibilities 7
AI 23O-001 Functional Importance Determination 0, 1, 2, 2A
AI 28A-001 Level 1 CR Evaluation (IIT) 12
AI 28A-003 Rapid Response To Events Of Significance 3A
AI 28A-006 Level 3 CR Evaluation 9
AI 28A-007 Level 2 CR Evaluation 4
AI 28A-010 Screening Condition Reports 6, 7
AI 28A-023 Evaluation of Maintenance Rule Functional Failure PIRs 1
AI 28E-007 Corrective Action Coding and Trend Analysis 7A
AI 28E-008 Condition Report Trend Code Assignment Process 2A
AI 34-003 Corrective Action Program Coordinators - Roles & 1A
Responsibilities
ALR KC-888 Fire Protection Panel KC-008 Alarm Response 18A
ALR 00-120A MFP A Trip 9, 10
ALR 00-120B MFP A Suct Press LO 11
ALR 00-016B PB03 / 04 Bus UV 11
AP 02-003 Chemistry Specification Manual 34B
AP 16C-006 MPAC Work Request/Work Order Process Controls 15
A1-2 Attachment 1
AP 18A-001 Employee Concerns Program 4
AP 20E-001 Industry Operating Experience Program 10, 11, 12,
13
AP 21-200 Operational Decision Making and Problem Analysis 3
AP 21C-001 WCGS/WESTAR Substation 8, 9, 11A
AP 22A-001 Screening, Prioritization and Pre-Approval 13
AP 22B-001 Outage Risk Management 9, 11, 12
AP 22C-003 Operational Risk Assessment Program 14A
AP 22C-003 On-Line Nuclear Safety and Generation Risk Assessment 15, 15A
AP 22C-005 IPS Daily Scheduling 12
AP 22C-007 Risk Management and Contingency Planning 4
AP 23M-001 WCGS Maintenance Rule Program 6
AP 23O-001 Plant Health Committee 4
AP 23-006 System Engineering Program 20, 20A
AP 23-008 Equipment Reliability Program 4
AP 24B-001 Control of Site Contractor Services 7
AP 26A-001 Reportable Events - Evaluation and Documentation 15
AP 26A-004 Communications With the Nuclear Regulatory Commission 10/11
AP 26A-007 NRC Performance Indicators 8
AP 28-011 Resolving Degraded or Nonconforming Conditions Impacting 2
AP 28A-100 Condition Reports 13
AP 36-001 Nuclear Safety Culture 0
CKL ZL-009 Site Reading Sheets 70
GEN 00-002 Cold Shutdown to Hot Standby 71
GEN 00-003 Hot Standby to Medium Load, pages 20 and 21 29
GEN 00-003, 10 Hot Standby to Medium Load 33
GEN 00-004 Power Operation, page 36 of 63 65
GEN 00-006 Hot Standby to Cold Shutdown 70
A1-3 Attachment 1
I-ENG-005 Infrared Thermograph 4
MPE E009Q-03 Inspection and Testing of Siemens Vacuum Circuit Breakers 4
OFN BB-031 Shutdown LOCA 17
OFN MA-038 Rapid Plant Shutdown, page 2 of 45 12
OFN NB-030 Loss of AC Emergency Bus NB01 (NB02) 20
OFN NB-035 Loss of Off-Site Power Restoration 0
RNM C-130 Miscellaneous Relay and Meter Equipment 6
STN AE-007 Startup Main Feedwater Pump Operational Test 0
SYS AE-132 MFIV Pressure Open with ASU (Laptop) 0 and 1
SYS AE-200 Operation of Feedwater Heating 7
STS AE-201 Feedwater Chemical Injection Inservice Valve Test 22
STS IC-215 TADOT of Manual Reactor Trip, Trip and Bypass Breaker 13, 14
UV/Shunt Trip, Turbine Trip on Reactor Trip and P-4
STS RE-018 Multiple Rod Drop Time Measurement 11
SYS AC-120 Main turbine Generator Startup 51
SYS AE-121 Turbine Driven Main Feedwater Pump Startup 32, 33
SYS AE-320 Turbine Driven Main Feedwater Pump Shutdown 23
SYS EJ-320 Placing RHR System in Safety Injection Standby Condition 34
SYS EJ-321 Shutdown of a Residual Heat Removal Train 29
SYS SB-122 Enabling/Disabling P-4/LO Tavg FWIS 1, 2
WCRE-13 Wolf Creek Generating Station Lake Water Systems 2
Structural Integrity Program
CONDITION REPORTS
777 7499 7508 7509 7510
7511 8575 9181 9375 9519
10247 10300 11768 12913 13805
13957 14261 14262 15269 15306
15407 15520 15521 15576 16455
16467 16657 16905 17776 17900
18034 18156 18413 19245 19295
A1-4 Attachment 1
19318 19360 19369 19371 19390
19447 19913 19914 19960 20665
21002 21039 21044 21260 21509
21641 21702 21813 21816 22470
22781 22979 23008 23032 23108
23110 23114 23119 23154 23479
23852 23938 23992 24445 24852
25018 25817 25892 25896 26345
26384 26787 26805 27005 27527
27997 27998 28175 28208 28224
28474 29095 29098 29128 29181
29204 29286 29818 30271 31121
31151 31458 31800 32326 32404
32431 32434 32436 32438 32445
32446 32451 32492 32506 33041
33076 33087 33103 33109 33143
33177 33202* 33212* 33217* 33229*
33253 33316* 33320 33327 33329*
33331* 33336* 33341* 33342* 33351*
33385* 33393* 33395* 33416* 33419*
33423* 33435* 33440* 33442* 33456*
33457* 33459* 33465* 33466* 33467*
33469* 33529* 33535* 33540* 33541*
33575* 33594* 33625* 33720* 33722*
33752* 33761* 33869* 33890* 33903
33917 33922* 33928* 33929* 33958*
33983* 33990* 2006-001409 2006-001798 2006-003035
2006-003271 2006-003473 2006-003815 2007-000040 2007-000187
2007-000202 2007-001013 2007-001707 2007-001780 2007-001993
2007-002000 2007-002009 2007-002128 2007-002331 2007-002749
2007-002854 2007-003350 2007-003612 2007-003798 2007-004055
2007-004125 2007-004126 2007-004127 2007-004128 2007-004129
2007-004130 2007-004132 2008-000116 2008-000149 2008-000164
2008-000465 2008-000989 2008-001014 2008-002230 2008-002237
A1-5 Attachment 1
2008-003419 2008-003802 2008-003810 2008-004136 2008-004536
2008-004997 2008-006105 PIR 1995-0586 PIR 1995-2858 PIR 1996-3260
PIR 1997-0078 PIR 1998-2794 PIR 2000-0834 PIR 2000-0835 PIR 2000-0871
PIR 2000-2212 PIR 2001-0041 PIR 2001-2368 PIR 2002-0860 PIR 2003-2178
PIR 2003-2496 PIR 2004-0586 PIR 2004-0684 PIR 2004-2435 PIR 2004-2502
PIR 2004-2644 PIR 2004-2684 PIR 2004-2813 PIR 2004-3390 PIR 2005-0121
PIR 2005-0382 PIR 2005-0771 PIR 2005-0773 PIR 2005-0774 PIR 2005-0775
PIR 2005-0776 PIR 2005-0777 PIR 2005-0778 PIR 2005-0779 PIR 2005-0780
PIR 2005-0781 PIR 2005-0782 PIR 2005-0783 PIR 2005-0784 PIR 2005-0785
PIR 2005-0786 PIR 2005-0787 PIR 2005-0788 PIR 2005-0789 PIR 2005-0790
PIR 2005-0791 PIR 2005-0792 PIR 2005-0794 PIR 2005-0795 PIR 2005-0796
PIR 2005-1411 PIR 2005-1962 PIR 2005-2126 PIR 2005-2164 PIR 2005-2167
PIR 2005-2168 PIR 2005-2461 PIR 2005-2507 PIR 2005-2619 PIR 2007-0483
- Condition Reports generated during the inspection
DRAWINGS
NUMBER TITLE REVISION
M-12AB03 Main Steam System 18
M-12AE01 Piping & Instrumentation Diagram Feedwater System 37
M-12AE02 Piping & Instrumentation Diagram Feedwater System 13
LERS
NUMBER TITLE DATE
2008-002-01 Technical Specification Allowed Outage Time Exceeded January 11, 2010
due to Room Cooler Leak
2008-003-00 Manual Reactor Trip due to Loss of Steam Generator May 13, 2008
Level
2008-004-02 Loss of Offsite Power Event When the Reactor was De- November 11,
fueled 2009
2008-007-00 Two Residual Heat Removal Trains Inoperable in July 10, 2007
Mode 3 due to Check Valve Leakage
2008-008-00 Potential for Residual Heat Removal Trains to be October 3, 2008
Inoperable during Mode Change
A1-6 Attachment 1
2008-008-02 Potential for Residual Heat Removal Trains to be August 25, 2009
Inoperable during Mode Change
2009-001-00 Reactor Protection System Actuation and Reactor Trip June 24, 2009
due to Main Feedwater Regulating Valve Failing Closed
2009-002-00 Loss of Offsite Power due to Lightning October 17, 2009
2009-005-00 Loss of Both Diesel Generators with all Fuel in the Spent December 21,
Fuel Pool 2009
2009-009-01 Defeating Feedwater Isolation on Low Tavg Coincident March 3, 2010
with P-4 Function Results in Missed Mode Change
2010-001-00 Automatic Start of Motor Driven Auxiliary Feedwater March 22, 2010
Pumps Inoperable During Startup in Mode 1
2010-002-00 Turbine Trip Function of Reactor Trip, P-4 Interlock March 29, 2010
Defeated During Entry into and in Mode 3
WORK ORDERS
00-223036-004 04-266765-000 05-273961-002 06-286463-000 07-291802-000
07-300375-000 08-303896-000 08-303897-000 09-316562-002 09-316566-004
09-316730-000 09-317186-000 09-317187-000 09-317188-000 09-317189-000
09-317190-000 09-317749-000 09-317750-000 09-317752-000 09-317753-000
09-317754-000 09-317755-000 09-317756-000 09-317757-000 09-317820-000
09-317821-000 09-317822-000 09-317823-000 09-321570-000 09-321571-000
09-321572-000 09-321573-000 09-322495-000 09-322503-000 10-324684-000
10-325088-000 10-325088-001 10-325205-000 10-326827-057 10-331761-000
10-332022-000 10-332233-000 10-332233-001 10-332631-000 10-332731-000
11-337163-002
MEETING NOTES
MEETING DATE
OE Collegial Review February 10, 2011
Level 4 Challenge Board February 23, 2011
Corrective Action Challenge Board February 24, 2011
A1-7 Attachment 1
MISCELLANEOUS
NUMBER TITLE DATE / REVISION
2010 Nuclear Safety Culture Assessment
Interview Responses
2010 Nuclear Safety Culture Assessment Pre-
assessment Survey
Amendment to Wolf Creek Generating Station August 13, 2010
Operating Agreement
Control Room Logs from April 7-8, 2008
Control Room Logs from November 17-20, 2008
Deficient/Corrective Maintenance Backlog January 2011
Reduction Initiative dated
Diesel Fuel Oil Strategic Plan 2
Employee Concern Program Overview 2009
Employee Concern Program Overview 2010
Engineering Technical Task Brief Desktop 1
Instruction
Essential Service Water Feasibility Study, Wolf November 2010
Creek Nuclear Operating Corporation, Burlington,
Kansas (WCNOC127)
FID Assignments for Fire Pumps, PN09, and
PB03
Feedwater Quick Hit Assessment November 12, 2010
GDC-17 Transmission Network Controls and 8
Reliability Improvement and Related Issues
HU Tools For Engineers Desktop Instruction 0
Human Performance Toolbox 3
Jan 2011 Management Observations of Ops
Letter ET 10-0011, dated March 4, 2010, from
T. J. Garrett WCNOC to NRC
List of PRA Top Ten Systems
Maintenance Rule Final Scope Evaluation
Essential Service Water (EF)
A1-8 Attachment 1
Metallurgical Failure Evaluation of a Corroded 30" November 25, 2009
Elbow from the Outlet Side of the Self-Cleaning
Strainer of an ESW Line, Report N0. 57809
Metallurgical Investigation of a Corroded 18" October 27, 2009
Welded Pipe, 150-HBC-18 from A ESW Lake
Water Line, Report No. 57652
MSDS for SPEC-AID 8Q5368ULS November 17, 2006
NERC Interface Coordination Agreement for the
Wolf Creek Substation between Westar Energy,
Inc and Wolf Creek Nuclear Operating
Corporation
Nuclear Plant Interface Coordination NUC 001-2 March 23, 2010
On the Spot Change (OTSC) 09-0084, Gen 00- November 17, 2009
002, Cold Shutdown to Hot
Standby On the Spot Change (OTSC) 09-0086, November 19, 2009
STS RE-018, Multiple Rod Drop Time
Measurement
On the Spot Change (OTSC) 10-0011, AP 22C- February 17, 2010
002, Work Controls
On the Spot Change (OTSC) 11-0014, February 22, 2011
Operational Focus Plan 1
Operational Focus Plan 2 (draft)
Operations Requalification Cycle 10-04 Schedule July 7, 2010
Preventative Maintenance Optimization (PMO) 0, 1, and 2
Project Report
Project Plan for Essential Service Water Piping January 18, 2011
Integrity Project 2
Project Report, GDC-17 Transmission Network 8
Controls and Reliability Improvement and Related
Issues
Quality Oversight Report December 2010/
January 2011
Quick Hit Detail Report 1684, PM Basis Review January 14, 2010
Quick Hit Detail Report 1795, STP STARS ER June 29, 2010
Benchmarking Trip
A1-9 Attachment 1
Record Supplement/Correction Sheet K01-033 for February 24, 2011
CR 23119 - Safety System Functional Failures
Exceeding PI Threshold
Single Point of Entry Training Presentations, September 2010
PowerPoint Slides
Slide presentation on Managing Defenses
Md 0510
Slide presentation on Reducing Error Re 0510
Student Notes from USA Event Reporting, 3A
RA270RPT.H02
Student Work Book for USA Event Reporting, 3A
RA270RPT.WB1
Today's Operational Focus February 7-10, 2011
Training Lesson Materials for General Training July 8, 2010
GT1535403, Rev 009, on Reportability - Event
Notification and Reporting
Training Lesson Slides for General Training 9
GT1535403, on Reportability - Event Notification
and Reporting
Training Roster dated Apr 21, 2010 for 15
AP 26A -001, Reportable Events - Evaluation and
Documentation
Training slides from USA Event Reporting The 3A
10 CFR 50.72 and 50.73, RA270RPT.PP1
Various Fuel Oil Lab Analysis Data Sheets
WCGS_PRA_Rev5_Raw Query
WCGS_PRA_Rev5_RRW Query
WCNOC Reportability Handbook 4
Wolf Creek Change Management Process Guide 1
Wolf Creek Change Plan Process High
Complexity Worksheet
Wolf Creek Change Plan Process Introduction
Worksheet
Wolf Creek Change Plan Process Low Complexity
Worksheet
A1-10 Attachment 1
Wolf Creek Change Plan Process Moderate
Complexity Worksheet
Wolf Creek Nuclear Operating Company Nuclear August 2008
Safety Culture Assessment
Wolf Creek Nuclear Operating Company Nuclear March 2010
Safety Culture Assessment
Wolf Creek Nuclear Operating Corporation, The April 13, 2010
Daily Current
Wolf Creek Nuclear Operating Corporation, The March 17, 2010
Daily Current
Wolf Creek Technical Specifications,
Amendment 188
Wolf Creek Updated Safety Analysis Report 23
(USAR)
00-223036-000 Tank Inspection Report 100K Diesel Fuel Oil April 2002
Tanks Alpha/Bravo Wolf Creek Nuclear Operating
Corporation (Vendor Report)
02846 Feedwater Preheating Calculation 1
AIF 22I-004-01 Project Ranking Points 2
APF 21C-001-01 WCGS Substation Work Authorization 5
Audit 09-02-ENG Quality Assurance Audit Report Engineering April 3, 2009
Program
Audit 10-02-OPS Quality Assurance Audit Report Operations March 16, 2010
Program
Audit 10-07-FP Quality Assurance Audit Report Fire Protection September 23, 2010
Program
Change Package Alarm 7300 Cabinets Card Frame Fuses 0
13343
CP 013043 Diesel Fuel Oil 0
FAQ 10-03 Wolf Creek Generating Station Unplanned March 18, 2010
Scrams with Complications
FSAR 10.4.7.2.3 Feedwater System Operation
OTSC 02-0108 50.59 Screening for GEN 00-003, Rev 54 April 27, 2002
procedure change
A1-11 Attachment 1
OTSC 02-0108 50.59 Applicability Determination for GEN 00-003, April 27, 2002
Rev 54 procedure change
RA1331201 Regulatory Awareness 5
RA2331201 Ai 26A-003 Other Regulatory Evaluations for Prior 0
NRC Approval
RA270RPT.LP1 Lesson Plan for USA Event Reporting 3A
SEL 2009-152 Self Assessment Report Predictive Maintenance
(PdM) Program
SEL 2010-194 Self Assessment Report Main Feedwater
SEL-2010-188 NRC Performance Indicator Program January 11, 2011
SEL 2010-192 INPO TR10-70 Self Assessment
SEL-2010-194 Main Feedwater Self Assessment December 7, 2010
SEL 01-033 Licensing Commitments
SEL 05-01 Transformer and Switchyard Self-Assessment
STN IC-903 Cross Trip Check XNB01 Switchyard 3
TB-05-6 Westinghouse Technical Bulletin - Retrofit of 0 and 1
Printed Circuit Cards for 7300 Based Systems -
Capacitor C105 Replacement with Fuse
Protection Added
TG1645500 Safety Culture - What's at Stake? 0
TIN ES1312300 Timeliness Evaluations RIS 2005-20 2
TSA 20273-000 Bechtel Response to Reactor Trips caused by 0 and 1
Main Feedwater
USN 153113 Feed Pump Turbine Upgrades Per TIL-1206 August 26, 2010
Recommendations
A1-12 Attachment 1