ML13226A255
| ML13226A255 | |
| Person / Time | |
|---|---|
| Site: | Wolf Creek  |
| Issue date: | 08/14/2013 |
| From: | O'Keefe N NRC/RGN-IV/DRP/RPB-B |
| To: | Matthew Sunseri Wolf Creek |
| References | |
| EA-13-084 IR-13-003 | |
| Download: ML13226A255 (90) | |
See also: IR 05000482/2013003
Text
UNITE D S TATE S
NUC LEAR RE GULATOR Y C OMMI S SI ON
R E G IO N I V
1600 EAST LAMAR BLVD
AR L I NGTON , TEXAS 7 60 11 - 4511
August 14, 2013
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 - INTEGRATED INSPECTION
REPORT NO. 05000482/2013003, NRC INVESTIGATION REPORT 4-2012-023,
AND NOTICE OF VIOLATION
Dear Mr. Sunseri:
On June 30, 2013, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at
the Wolf Creek Generating Station. In addition, the NRC Office of Investigations, Region IV
completed an investigation on March 28, 2013. The purpose of the investigation was to
determine whether an individual, formerly employed by Wolf Creek Generating Station, falsified
procedure paperwork. The enclosed inspection report documents the inspection results which
were discussed on July 11, 2013, with Mr. J. Broschak, Vice President of Engineering, and
other members of your staff. A supplement exit was conducted on August 7, 2013, with Mr. R.
Smith, Site Vice President and Chief Nuclear Operations Officer.
The inspections examined activities conducted under your license as they relate to safety and
compliance with the Commission's rules and regulations and with the conditions of your license.
The inspectors reviewed selected procedures and records, observed activities, and interviewed
personnel.
Based on the results of this inspection and the information developed during the investigation,
the NRC has determined that a violation of NRC requirements occurred (EA-13-084). The
violation is cited in the enclosed Notice of Violation (Notice) and the circumstances surrounding
it are described in detail in the subject inspection report. Because the violation is associated
with willfulness, it was evaluated under the traditional enforcement process as set forth in the
NRC Enforcement Policy. The NRC concluded that the violation, absent willfulness, would be
considered a minor violation because the failure to complete and document the inspection per
the procedure did not have any safety significance.
However, the NRC considers the violation to have been more significant than minor, because it
involved willfulness, and therefore, the NRC has classified the violation at Severity Level IV, in
accordance with the NRC Enforcement Policy. The current Enforcement Policy is included on
the NRC's Web site at (http://www.nrc.gov/about-nrc/regulatory/enforcement/enforce-pol.html).
M. Sunseri -2-
You are required to respond to this letter and should follow the instructions specified in the
enclosed Notice when preparing your response. If you have additional information that you
believe the NRC should consider, you may provide it in your response to the Notice. The NRC
review of your response to the Notice will also determine whether further enforcement action is
necessary to ensure compliance with regulatory requirements.
In addition, three NRC identified and two self-revealing findings of very low safety significance
(Green) were identified during this inspection. Each of these findings was determined to involve
violations of NRC requirements. Additionally, the NRC has determined that a traditional
enforcement Severity Level IV violation occurred. This traditional enforcement violation was
identified without an associated finding. The NRC is treating the NRC identified and self-
revealing findings as non-cited violations (NCVs), consistent with Section 2.3.2 of the
Enforcement Policy. These NCVs are described in the subject inspection report.
If you contest the violation or the significance of these NVCs, 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 Wolf Creek Generating Station.
If you disagree with a cross-cutting aspect assignment in this report, you should provide a
response within 30 days of the date of this inspection report, with the basis for your
disagreement, to the Regional Administrator, Region IV; and the NRC Resident Inspector at
Wolf Creek Generating Station.
In accordance with 10 CFR 2.390 of the NRC's "Rules of Practice," a copy of this letter, its
enclosures, and your response, if you choose to provide one, will be available electronically for
public inspection in the NRC Public Document Room or from the Publicly Available
Records (PARS) component of NRC's Agencywide Document Access and Management
System (ADAMS). ADAMS is accessible from the NRC Web site at http://www.nrc.gov/reading-
rm/adams.html (the Public Electronic Reading Room). To the extent possible, your response
should not include any personal privacy or proprietary, information so that it can be made
available to the Public without redaction.
Sincerely,
/RA/
Neil O'Keefe, Chief
Project Branch B
Division of Reactor Projects
Docket No.: 50-482
License No: NPF-42
M. Sunseri -3-
Enclosures:
1. Notice of Violation
2. Inspection Report 05000482/2013003
w/Attachments:
1. Supplemental Information
2. Information Request for Inspection Activities, documented in 71111.08
3. Information Request for Inspection Activities, documented in 71124.01
cc w/encl:
Electronic Distribution for Wolf Creek Generating Station
SUNSI Rev Compl. Yes No ADAMS Yes No Reviewer Initials NFO
Publicly Avail. Yes No Sensitive Yes No Sens. Type Initials NFO
SRI:DRP/B RI:DRP/B SPE:DRP/B C:DRS/TSB C:DRS/EB1 C:DRS/EB2
CPeabody CHunt MBloodgood RKellar TFarnholtz GMiller
/RA/TFarnholtz
/RA/E /RA/E /RA/ /RA/ /RA/
for
08/05/13 08/12/13 08/13/13 08/07/13 08/07/13 08/09/13
C:DRS/OB C:DRS/PSB1 C:DRS/PSB2 C:ORA/ACES RC BC:DRP/B
VGaddy MHaire JDrake HGepford KFuller NOkeefe
/RA/for /RA/ /RA/ /RA/ /RA/E /RA/
08/07/13 08/07/13 08/08/13 08/13/13 08/08/13 08/13/13
NOTICE OF VIOLATION
Wolf Creek Nuclear Operating Corporation Docket No. 50-482
Wolf Creek Generating Station License No. NFP-42
During an NRC inspection conducted on June 30, 2013, and an NRC investigation completed
on March 28, 2013, a violation of NRC requirements was identified. In accordance with the
NRC Enforcement Policy, the violation is listed below:
10 CFR 50.9 requires, in part, that information required by statute, orders, or license
conditions to be maintained by the licensee shall be complete and accurate in all
material respects.
Wolf Creek License Condition 2.C.5, Fire Protection, requires that the licensee shall
maintain in effect all provisions of the approved fire protection program as described in
the Standardized Nuclear Unit Power Plant System Final Safety Analysis Report.
Section 9.5-1 of the Wolf Creek Updated Final Safety Analysis Report, dated
March 10, 2013, describes the fire protection program and includes the licensees
commitment to meet Appendix 3A, Conformance to NRC Regulatory Guides, and
Appendix A, Table 9.5A-1 of Regulatory Guide1.39, Housekeeping Requirements for
Water-Cooled Nuclear Power Plants, Revision 2. Regulatory Guide 1.39, Revision 2,
endorses ANSI Standard N45.2.3-1973, Housekeeping During the Construction Phase
of Nuclear Power Plants.
ANSI Standard N45.2.3-1973, Section 3.5, states, in part, that periodic inspection and
examination of the work areas shall be performed at scheduled intervals to assure
adequacy of cleanliness and housekeeping practices. Section 4 of the above ANSI
Standard states, in part, that copies of inspection and examination records shall be
prepared and placed with other project records.
Section 6.1.8 of Procedure AP 12-001, Housekeeping Control, Revisions 6C and 7,
dated May 5, 2006, and November 10, 2008, respectively, intended to implement the
inspection and examination requirements of ANSI Standard N45.2.3-1973, states, in part
that assigned personnel shall walk down their areas monthly and that personnel
record and document their walkdowns using the Housekeeping Inspection Card.
Contrary to the above, between October and December 2008, the licensee failed to
maintain records required by License Condition 2.C.5 that were complete and accurate
in all material respects. Specifically, the Housekeeping Inspection Card for the spent
fuel pool area indicated that the inspection had been completed by a certain individual.
Security access logs, however indicated that the individual that completed the record
(Housekeeping Inspection Card) had not entered the area. This information is material
because it provides assurance to the NRC that the licensee has performed periodic
inspection and examination of work areas at scheduled intervals to assure adequacy of
cleanliness and housekeeping practices as required by the license condition.
-1- Enclosure 1
This is a Severity Level IV violation. (Section 6.9)
Pursuant to the provisions of 10 CFR 2.201, Wolf Creek Nuclear Operating Corporation is
hereby required to submit a written statement or explanation to the U.S. Nuclear Regulatory
Commission, ATTN: Document Control Desk, Washington, DC 20555-0001 with a copy to the
Regional Administrator, Region IV, and a copy to the NRC Resident Inspector at the Wolf Creek
facility within 30 days of the date of the letter transmitting this Notice of Violation (Notice). This
reply should be clearly marked as a "Reply to a Notice of Violation; EA-13-084" and should
include for each violation: (1) the reason for the violation, or, if contested, the basis for disputing
the violation or severity level, (2) the corrective steps that have been taken and the results
achieved, (3) the corrective steps that will be taken, and (4) the date when full compliance will
be achieved. Your response may reference or include previous docketed correspondence, if
the correspondence adequately addresses the required response. If an adequate reply is not
received within the time specified in this Notice, an order or a Demand for Information may be
issued as to why the license should not be modified, suspended, or revoked, or why such other
action as may be proper should not be taken. Where good cause is shown, consideration will
be given to extending the response time.
If you contest this enforcement action, you should also provide a copy of your response, with
the basis for your denial, to the Director, Office of Enforcement, United States Nuclear
Regulatory Commission, Washington, DC 20555-0001.
Because your response will be made available electronically for public inspection in the NRC
Public Document Room or from the NRCs Agencywide Documents and Access Management
System (ADAMS), accessible from the NRC Web site at http://www.nrc.gov/reading-
rm/adams.html, to the extent possible, it should not include any personal privacy, proprietary, or
safeguards information so that it can be made available to the public without redaction. If
personal privacy or proprietary information is necessary to provide an acceptable response,
then please provide a bracketed copy of your response that identifies the information that
should be protected and a redacted copy of your response that deletes such information. If you
request withholding of such material, you must specifically identify the portions of your response
that you seek to have withheld and provide in detail the bases for your claim of withholding (e.g.,
explain why the disclosure of information will create an unwarranted invasion of personal
privacy or provide the information required by 10 CFR 2.390(b) to support a request for
withholding confidential commercial or financial information). If safeguards information is
necessary to provide an acceptable response, please provide the level of protection described
in 10 CFR 73.21.
In accordance with 10 CFR 19.11, you may be required to post this Notice within two working
days of receipt.
Dated this 13th day of August, 2013
-2- Enclosure 2
U.S. NUCLEAR REGULATORY COMMISSION
REGION IV
Docket: 05000482
License: NPF-42
Report: 05000482/2013003
Licensee: Wolf Creek Nuclear Operating Corporation
Facility: Wolf Creek Generating Station
Location: 1550 Oxen Lane NE, Burlington, Kansas
Dates: March 31 through June 30, 2013
Inspectors: C. Peabody, Senior Resident Inspector
C. Hunt, Acting Resident Inspector
M. Bloodgood, Senior Project Engineer
R. Kopriva, Senior Reactor Inspector
L. Ricketson P.E., Senior Health Physicist
B. Correll, Reactor Inspector
J. ODonnell, Health Physicist
C. Speer, Reactor Inspector
M. Williams, Reactor Inspector
Approved By: Neil O'Keefe, Chief, Project Branch B
Division of Reactor Projects
-1- Enclosure 2
SUMMARY OF FINDINGS
IR 05000482/2013003, 03/31 - 06/30/2013, Wolf Creek Generating Station, Integrated Resident
and Regional Report; Inservice Inspection Activities, Follow-up of Events and Notices of
Enforcement Discretion, Other Activities.
The report covered a 3-month period of inspection by resident inspectors and announced
baseline inspections by region-based inspectors. Five Green non-cited violations of significance
were identified. One Severity Level IV 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 cross-cutting 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.
A. NRC-Identified Findings and Self-Revealing Findings
Cornerstone: Initiating Events
- Green. The inspectors identified a Green non-cited violation of 10 CFR Part 50,
Appendix B, Criterion V, Instructions, Procedures, and Drawings, which states,
in part, activities affecting quality shall be prescribed by procedures of a type
appropriate to the circumstances and accomplished in accordance with these
procedures. Contrary to the above, the licensee failed to ensure procedures
related to the boric acid corrosion control program were adequate and properly
implemented. Specifically, prior to February 19, 2013, the licensee failed to:
(1) resolve discrepancies within the boric acid corrosion control program
procedure; (2) resolve discrepancies between the boric acid corrosion control
program procedure and the boric acid leak management procedure; and
(3) failed to track and resolve leakage for locations where health physics had
installed drip catch containments, to review the Health Physics Drip Bag Log as
part of the quarterly outside containment walkdown, and to add component
locations to the program. Further, the licensee failed to periodically assess the
effectiveness of the program on a refueling frequency. The violation was entered
into the licensees corrective action program as Condition Report 65212.
The inspectors determined that the failure to recognize discrepancies between
boric acid control procedures and the failure to follow boric acid program
procedures was a performance deficiency. The performance deficiency was
more than minor because it affected the Initiating Events Cornerstone attribute of
procedure quality and affected the cornerstone objective to limit the likelihood of
those events that upset plant stability and challenge critical safety functions
during shutdown as well as power operations, and if left uncorrected, the
performance deficiency had the potential to lead to a more significant safety
concern. Specifically, failure to resolve discrepancies within procedures or track
-2- Enclosure 2
and resolve leak locations where health physics had installed drip catch
containments had the potential to mischaracterize leaks or allow leaks to corrode
safety-related systems. Using Inspection Manual Chapter 0609, Appendix A,
The Significance Determination Process for Findings At-Power, the finding was
determined to be of very low safety significance (Green), because the finding
was a procedure quality problem that did not represent a loss of system safety
function, and did not screen as potentially risk significant due to a seismic,
flooding, or severe weather initiating event. The finding had a cross-cutting
aspect in the area of human performance associated with the work practices
component because the licensee failed to ensure supervisory and management
oversight of work activities, including procedure appropriateness and compliance,
such that nuclear safety is supported H.4(c) (Section 1R08.3.b.1).
- Green. The inspectors identified a Green non-cited violation of 10 CFR Part 50,
Appendix B, Criterion XVI, Corrective Action, which states, in part, Measures
shall be established to assure that conditions adverse to quality are promptly
identified and corrected. Contrary to the above, the licensee failed to identify
and correct a condition adverse to quality in a timely manner. Specifically, prior
to February 19, 2013, the licensee failed to document the large area of boric acid
leakage and corroded steel plates on the south primary shield wall of the
containment refueling pool. The violation was entered into the licensees
corrective action program as Condition Report 64213.
The inspectors determined that the failure to promptly identify and evaluate a
condition adverse to quality was a performance deficiency. The performance
deficiency was more than minor because it affected the Initiating Events
Cornerstone attribute of human performance and affected the cornerstone
objective to limit the likelihood of those events that upset plant stability and
challenge critical safety functions during shutdown as well as power operations,
and if left uncorrected, the performance deficiency had the potential to lead to a
more significant safety concern. Specifically, failure to implement corrective
actions could result in increased leakage and further degradation of the safety
system. Using Manual Chapter 0609.04, Phase 1 - Initial Screening and
Characterization of Findings, the inspectors determined that this finding was of
very low safety significance (Green), because it was not a design or qualification
deficiency, did not represent a loss of system safety function, and did not screen
as potentially risk significant due to a seismic, flooding, or severe
weather initiating event. The finding had a cross-cutting aspect in the area of
human performance associated with the work practices component because the
licensee failed to define and effectively communicate expectations regarding
procedural compliance and that personnel follow procedures H.4(b)
(Section 1R08.3.b.2).
- Green. A Green self-revealing non-cited violation of Technical Specification
5.4.1.a was identified for failure to properly update operating procedures and
train operators on the effects of a recently installed modification. Specifically,
procedures were not adequately revised to provide guidance for operating the
-3- Enclosure 2
new Westinghouse Ovation digital turbine controls. As a result, operators shifted
operating modes at a power level that caused an 11 percent power increase due
to the combined characteristics of the steam control valves and the turbine
control unit. Additionally, operators were trained to shift control modes at low
power levels, where minor transients occurred, but were not restricted from
performing the shift at high power levels, where the transient could be more
significant. This issue was entered into the licensees corrective action program
under Condition Report 68711.
Failure to update station operating procedures to provide adequate guidance for
design changes, and failure to adequately train operators on those implemented
design changes is a performance deficiency. The performance deficiency is
more than minor because it affected the design control, procedure quality, and
human performance attributes of the Initiating Events cornerstone objective to
limit the likelihood of events that upset plant stability and challenge critical safety
functions during shutdown as well as power operations. Using Inspection
Manual Chapter 0609, Appendix A, Checklist 1, Initiating Events Screening
Questions, the inspectors determined that the finding was of very low safety
significance (Green) because the finding did not result in a reactor trip coincident
with the loss of mitigation equipment. The inspectors determined that this finding
had a cross-cutting aspect in the area of human performance area of work
control, because the licensee did not appropriately communicate and coordinate
during activities in which interdepartmental coordination was necessary to assure
plant and human performance. Specifically, Wolf Creek did not communicate
and coordinate to ensure that procedure guidance and operator training
adequately conveyed the operational impacts of shifting turnine control modes at
different power levels H.3(b) (Section 4OA3.5.b.1).
- Green. Inspectors identified a Green non-cited violation of Technical
Specification 5.4.1.a for the failure to follow Conduct of Operations and Reactivity
Management procedures. The inspectors reviewed an unplanned 11 percent
power increase during a shift in turbine control modes, and identified that pre-job
briefings did not adequately discuss expected plant response, operators did not
take action to limit the power increase when an unexpected response was
observed, and management was not adequately involved in decision making
prior to continuing power ascension before the details of an apparent turbine
control malfunction were fully understood. This issue was entered into the
licensees corrective action program under Condition Report 68711.
Failure to provide contingency actions for a greater than anticipated reactor
transient in the pre-job reactivity brief, and continuing with power ascension
without understanding the cause of the unexpected turbine control system
behavior is a performance deficiency. The performance deficiency is more than
minor because it affected the human performance attributes of the Initiating
Events cornerstone objective to limit the likelihood of events that upset plant
stability and challenge critical safety functions during shutdown as well as power
operations. Using Inspection Manual Chapter 0609 Appendix A, Checklist 1,
-4- Enclosure 2
Initiating Events Screening Questions, and the inspectors determined that the
finding was of very low safety significance (Green) because the finding did not
result in a reactor trip coincident with the loss of mitigation equipment. The
inspectors determined that this finding had a cross-cutting aspect in the area of
human performance area of work practices because the licensee failed to
communicate human error prevention techniques, such as holding pre-job
briefings, self and peer checking, and proper documentation of activities such
that work activities were performed safely. In addition, personnel proceeded in
the face of uncertainty or unexpected circumstances. Specifically, in the first
example control room operators pre-job reactivity brief was not appropriate
commensurate with the risk of the assigned task; in the second example station
personnel proceeded in the face of uncertainty H.4(a) (Section 4OA3.5.b.2).
Cornerstone: Mitigating Systems
- Green. A self-revealing non-cited violation of 10 CFR Part 50 Appendix B,
Criterion XVI, Corrective Action, was identified on March 13, 2013. Specifically,
the licensee replaced a jacket water pressure transmitter ten times, but failed to
correct pressure oscillations that caused a fatigue failure of a pressure switch
diaphragm, which rendered emergency diesel generator B inoperable. The
inspectors concluded that the licensee ineffectively focused on correcting the
apparent source of the pressure oscillations, but failed to evaluate the effects of
the pressure cycles on components exposed to the same oscillations. This issue
was entered into the licensees corrective action program as Condition
Report 65624.
Failure to analyze the effects of pressure oscillations in the emergency diesel
jacket water system on interfacing system components is a performance
deficiency. The performance deficiency is more than minor because it affected
the equipment performance attribute of the Mitigating Systems cornerstone
objective to ensure the availability, reliability, and capability of systems that
respond to initiating events to prevent undesirable consequences. Using
Inspection Manual Chapter 0609 Appendix A, Significance Determination
Process for Findings At Power, and determined that the finding screens as very
low safety significance (Green) because the finding does not meet any criteria
outlined in the Exhibit 2, Section A. Specifically the finding did not represent a
loss of system safety function and did not exceed its technical specification
allowed outage time of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. The inspectors determined that the finding had
a cross-cutting aspect in the area of problem identification and resolution
evaluations because the licensee failed to ensure that issues that potentially
affect nuclear safety are fully evaluated and addressed in a timely manner
P.1(c) (Section 4OA3.3).
-5- Enclosure 2
Cornerstone: Barrier Integrity
- SLIV. The inspectors identified a Severity Level IV violation of 10 CFR 50.9,
Completeness and accuracy of information, for the Wolf Creek Nuclear
Generating Stations failure to maintain complete and accurate records required
by a license condition. Title 10 CFR 50.9 requires, in part, that information
required by statute, orders, or license conditions to be maintained by the licensee
shall be complete and accurate in all material respects. Contrary to the above,
between October and December 2008, the licensee failed to maintain records
required by License Condition 2.C.5 that were complete and accurate in all
material respects. Specifically, the Housekeeping Inspection Card for the spent
fuel pool area indicated that the inspection had been completed when security
access logs indicate that the individual that completed the record had not entered
the area. The NRC investigation determined that the assigned individual did not
walk down the assigned area, and did not assign a designee to do so
(EA-13-084).
The failure to maintain records required by License Condition that are complete
and accurate in all material respects in accordance with 10 CFR 50.9 was a
violation. Because the violation is associated with willfulness and impacted the
regulatory process it was evaluated under the traditional enforcement process as
set forth in the NRC Enforcement Policy. Since this violation was the result of a
willful action, the NRC considers the violation to be more than minor, and
therefore, the NRC has classified the violation at Severity Level IV, in accordance
with the NRC Enforcement Policy (Section 4OA5).
B. Licensee-Identified Violations
None
-6- Enclosure 2
PLANT STATUS
The inspection period began with the unit in Mode 5 (cold shutdown) coming back from a
refueling outage in progress. The plant started up on April 13, 2013, and reached 100 percent
power on April 19, 2013. On April 29, 2013, the unit reduced power and the turbine was taken
off line to repair a stator cooling water leak. The turbine was restarted on May 2, 2013, the
same day the reactor experienced an unplanned transient (11percent power increase) while
shifting operating modes in the turbine steam controls. The unit returned to 100 percent power
on May 3, 2013. On May 7, 2013, the unit conducted a technical specification-required
shutdown due to a non-functional Class 1E air conditioner, and achieved cold shut down for
repairs the following day. The unit was restarted on May 13, 2013, and reached 100 percent
power on May 15. On June 17, 2013, the recently repaired Class 1E air conditioner showed
signs of substantial internal degradation. The unit began a technical specification-required
shutdown, reaching 16 percent power before the licensee was granted a Notice of Enforcement
Discretion to allow 7 days to replace the air conditioner again. Power was restored to
100 percent on June 19, 2013.
REPORT DETAILS
1. REACTOR SAFETY
Cornerstones: Initiating Events, Mitigating Systems, and Barrier Integrity
1R01 Adverse Weather Protection (71111.01)
.1 Readiness for Impending Adverse Weather Conditions
a. Inspection Scope
Since thunderstorms with potential tornados and high winds were forecast in the vicinity
of the facility for June 27, 2013, the inspectors reviewed the plant personnels overall
protection for the expected weather conditions. On June 28, 2013, the inspectors
walked down the main transformer system because the oil cooling pumps circuit breaker
had tripped in the previous nights electrical storm, possibly due to lightning in the
vicinity. The inspectors evaluated the plant staffs recovery actions against the sites
procedures to verify whether the staffs actions were adequate. During the inspection,
the inspectors focused on plant-specific design features and the licensees procedures
used to respond to specified adverse weather conditions. The inspectors also toured the
plant grounds to look for any loose debris that could become missiles during a
subsequent storm. The inspectors evaluated operator staffing and accessibility of
controls and indications for those systems required to control the plant. Additionally, the
inspectors reviewed the Updated Safety Analysis Report and performance requirements
for the systems selected for inspection, and verified that operator actions were
appropriate as specified by plant-specific procedures. The inspectors also reviewed a
sample of corrective action program items to verify that the licensee-identified adverse
weather issues at an appropriate threshold and dispositioned them through the
corrective action program in accordance with station corrective action procedures.
Specific documents reviewed during this inspection are listed in the attachment.
-7- Enclosure 2
These activities constitute completion of one adverse weather sample as defined in
Inspection Procedure 71111.01.
b. Findings
No findings were identified.
1R04 Equipment Alignment (71111.04)
.1 Partial Walkdown
a. Inspection Scope
The inspectors performed partial system walkdowns of the following risk-significant
systems:
- May 21, 2013, essential service water train B
- June 5, 2013, motor-driven auxiliary feedwater train A
- June 5, 2013, motor-driven auxiliary feedwater train B
The inspectors selected these systems based on their risk significance relative to the
reactor safety cornerstones at the time they were inspected. The inspectors attempted
to identify any discrepancies that could affect the function of the system and, therefore,
potentially increase risk. The inspectors reviewed applicable operating procedures,
system diagrams, Updated Safety Analysis Report, technical specification requirements,
administrative technical specifications, outstanding work orders, condition reports, and
the impact of ongoing work activities on redundant trains of equipment in order to identify
conditions that could have rendered the systems incapable of performing their intended
functions. The inspectors also inspected accessible portions of the systems to verify
system components and support equipment were aligned correctly and operable. The
inspectors examined the material condition of the components and observed operating
parameters of equipment to verify that there were no obvious deficiencies. The
inspectors also verified that the licensee had properly identified and resolved equipment
alignment problems that could cause initiating events or impact the capability of
mitigating systems or barriers and entered them into the corrective action program with
the appropriate significance characterization. Specific documents reviewed during this
inspection are listed in the attachment.
These activities constitute completion of three partial system walkdown samples as
defined in Inspection Procedure 71111.04-05.
b. Findings
No findings were identified.
-8- Enclosure 2
.2 Complete Walkdown
a. Inspection Scope
On April 30, 2013, the inspectors performed a complete system alignment inspection of
the diesel generator B starting air system to verify the functional capability of the system.
The inspectors selected this system because it was considered both safety significant
and risk significant in the licensees probabilistic risk assessment. The inspectors
inspected the system to review mechanical and electrical equipment lineups, electrical
power availability, system pressure and temperature indications, as appropriate,
component labeling, component lubrication, component and equipment cooling, hangers
and supports, operability of support systems, and to ensure that ancillary equipment or
debris did not interfere with equipment operation. The inspectors reviewed a sample of
past and outstanding work orders to determine whether any deficiencies significantly
affected the system function. In addition, the inspectors reviewed the corrective action
program database to ensure that system equipment-alignment problems were being
identified and appropriately resolved. Specific documents reviewed during this
inspection are listed in the attachment.
These activities constitute completion of one complete system walkdown sample as
defined in Inspection Procedure 71111.04-05.
b. Findings
No findings were identified.
1R05 Fire Protection (71111.05)
.1 Quarterly Fire Inspection Tours
a. Inspection Scope
The inspectors conducted fire protection walkdowns that were focused on availability,
accessibility, and the condition of firefighting equipment in the following risk-significant
plant areas:
- April 24, 2013, diesel generator B room, fire area D-1
- May 20, 2013, emergency core cooling systems B train (safety injection,
centrifugal charging, and containment spray pump rooms), fire area A-4
- May 20, 2013, Class 1E 4kV switchgear B room, fire area C-10
- May 28, 2013, control room air conditioning room B, fire area A-21
- May 28, 2013, control room air conditioning room A, fire area A-22
-9- Enclosure 2
The inspectors reviewed areas to assess if licensee personnel had implemented a fire
protection program that adequately controlled combustibles and ignition sources within
the plant; effectively maintained fire detection and suppression capability; maintained
passive fire protection features in good material condition; and had implemented
adequate compensatory measures for out of service, degraded or inoperable fire
protection equipment, systems, or features, in accordance with the licensees fire plan.
The inspectors selected fire areas based on their overall contribution to internal fire risk
as documented in the plants Individual Plant Examination of External Events with later
additional insights, their potential to affect equipment that could initiate or mitigate a
plant transient, or their impact on the plants ability to respond to a security event. Using
the documents listed in the attachment, the inspectors verified that fire hoses and
extinguishers were in their designated locations and available for immediate use; that
fire detectors and sprinklers were unobstructed; that transient material loading was
within the analyzed limits; and fire doors, dampers, and penetration seals appeared to
be in satisfactory condition. The inspectors also verified that minor issues identified
during the inspection were entered into the licensees corrective action program.
Specific documents reviewed during this inspection are listed in the attachment.
These activities constitute completion of five quarterly fire-protection inspection samples
as defined in Inspection Procedure 71111.05-05.
b. Findings
No findings were identified.
.2 Annual Fire Protection Drill Observation (71111.05A)
a. Inspection Scope
On April 11, 2013, the inspectors observed a fire brigade activation response to an
actual fire near the auxiliary boiler exhaust stack. The observation evaluated the
readiness of the plant fire brigade to fight fires. The inspectors verified that the licensee
staff identified deficiencies, openly discussed them in a self-critical manner, and took
appropriate corrective actions. Specific attributes evaluated were (1) proper wearing of
turnout gear and self-contained breathing apparatus; (2) proper use and layout of fire
hoses; (3) employment of appropriate fire fighting techniques; (4) sufficient firefighting
equipment brought to the scene; (5) effectiveness of fire brigade leader communications,
command, and control; (6) search for victims and propagation of the fire into other plant
areas; and (7) utilization of preplanned strategies.
These activities constitute completion of one annual fire-protection inspection sample as
defined in Inspection Procedure 71111.05-05.
b. Findings
No findings were identified.
- 10 - Enclosure 2
1R07 Heat Sink Performance (71111.07)
a. Inspection Scope
The inspectors reviewed licensee programs, verified performance against industry
standards, and reviewed critical operating parameters and maintenance records for the
essential service water (ESW)/service water macro foul treatment on June 12, 2013.
The inspectors verified that performance tests were satisfactorily conducted for heat
exchangers/heat sinks and reviewed for problems or errors; the licensee utilized the
periodic maintenance method outlined in Electric Power Research Institute (EPRI)
Report NP 7552, Heat Exchanger Performance Monitoring Guidelines; the licensee
properly utilized biofouling controls; the licensees heat exchanger inspections
adequately assessed the state of cleanliness of their tubes; and the heat exchanger was
correctly categorized under 10 CFR 50.65, Requirements for Monitoring the
Effectiveness of Maintenance at Nuclear Power Plants. Specific documents reviewed
during this inspection are listed in the attachment.
These activities constitute completion of one annual heat sink inspection sample as
defined in Inspection Procedure 71111.07-05.
b. Findings
No findings were identified.
1R08 Inservice Inspection Activities (71111.08)
Completion of Sections .1 through .5, below, constitutes completion of one sample as
defined in Inspection Procedure 71111.08-05.
.1 Inspection Activities Other Than Steam Generator Tube Inspection, Pressurized Water
Reactor Vessel Upper Head Penetration Inspections, and Boric Acid Corrosion Control
(71111.08-02.01)
a. Inspection Scope
The inspectors observed 16 nondestructive examination activities and reviewed
two nondestructive examination packages that included five types of examinations.
The inspectors directly observed the following nondestructive examinations:
SYSTEM WELD IDENTIFICATION EXAMINATION TYPE
Main Steam Report # MT3954: Feedwater heater Magnetic Particle
flange, Work Package # 13-364385-000,
Drawing # M-010A-0054
Essential Service Report # MT3964, Work Order 11- Magnetic Particle
Water 341145-000, Drawing EFV0062, RHR
- 11 - Enclosure 2
SYSTEM WELD IDENTIFICATION EXAMINATION TYPE
pump room B, ESW valve EFV00062
Essential Service Report # MT3963, Work Order 11- Magnetic Particle
Water 340718-002, Drawing EFV0061, RHR
pump room B, ESW valve EFV00061
Reactor Coolant Report # 3705, Work Order 11-339304- Penetrant
System 005, Weld ID# W1A
Safety relief valve drain line valve
BBV0088
Reactor Coolant Report # 3606, Work Order 11-339304- Penetrant
System 005, Drawing # M-13BB13
Safety relief valve drain line valve
BBV0085
Reactor Coolant Report # 4236- Work Order 11-339304- Radiograph
System 005, ID # W-2A, Reactor pressurizer
safety relief valve drain line valve
BBV0088
Reactor Coolant Report # 4235, Work Order 11-339208- Radiograph
System 006, ID # W-3A , Reactor pressurizer
safety relief valve drain line valve
BBV0085
Reactor Coolant Report # RF19 JEW-004. Longitudinal Ultrasonic
System seam weld on reactor pressurizer (shell
to shell weld). ISI # TBB03-SEAM-1-W
Reactor Coolant Report # RF19-JLD-001. Examination of Ultrasonic
System weld overlay on reactor pressurizer
surge line (including both the nozzle to
safe-end dissimilar metal weld and the
safe-end to pipe stainless steel weld).
ISI #TBB03-MW7090-WOL-DM and #
TBB03-MW7090-WOL-SS.
Reactor Coolant Report # RF19 GPF-004. Reactor Ultrasonic
System pressurizer surge nozzle inner radius
examination. ISI # TBB03-10A-IR.
Reactor Coolant Report # RF19-JEW-005. Reactor Ultrasonic
System pressurizer surge nozzle to shell weld.
ISI # TBB03-10A-W
Main Steam Report # RF19-GPF-005. 28 inch Fluted Ultrasonic
Head to Pipe. ISI # AB-01-F050, Loop 3
Circumferential Weld.
- 12 - Enclosure 2
SYSTEM WELD IDENTIFICATION EXAMINATION TYPE
Reactor Coolant Work Order 339304-005, ID # W-1A and Visual
System W-4A, Safety relief valve drain line valve
BBV0088
Reactor Coolant Work Order 339304-000, ID # W-1B Visual
System and W-4A, Safety relief valve drain line
valve BBV0085
Main Steam Work Order 11-344165-005 Visual
ID # AB-01-C011, Room 1412 Area 5,
Support and hanger
Main Steam Work Order 11-344165-005 Visual
ID # AB-01-H005, Room 1412 Area 5,
Piping Support
The inspectors reviewed records for the following nondestructive examinations:
SYSTEM WELD IDENTIFICATION EXAMINATION TYPE
Reactor Coolant Work Order 339304-000, ID # W-2 and Visual
System W-3A, Safety relief valve drain line valve
BBV0085
Reactor Coolant Work Order 339304-005, ID # W-2B and Visual
System W-3, Safety relief valve drain line valve
BBV0088
During the review and observation of each examination, the inspectors verified that
activities were performed in accordance with the American Society of Mechanical
Engineers (ASME) Code requirements and applicable procedures. There were no
relevant conditions identified for ASME Code Class 1 and 2 systems since the beginning
of the last refueling outage. The inspectors also verified that the qualifications of
nondestructive examination technicians performing the inspections were current.
The inspector observed the following welding activities:
SYSTEM WELD IDENTIFICATION WELD TYPE
Reactor Coolant Work Order 339304-000, ID # W- Tungsten Inert Gas
System 1B and W-4A, Safety relief valve Welding (GTAW)
drain line valve BBV0085
Reactor Coolant Work Order 339304-005, ID # W- Tungsten Inert Gas
System 1A and W-4A, Safety relief valve Welding (GTAW)
drain line valve BBV0088
- 13 - Enclosure 2
Essential Service RHR pump room, ESW room Tungsten Inert Gas
Water cooler valve replacement. Valve Welding (GTAW)
EFV0061, Work Order 11-34018-
002, Drawing # M-13EF04, ID #
PW-1A and PW-2A
Essential Service RHR pump room, ESW room Tungsten Inert Gas
Water cooler valve replacement. Valve Welding (GTAW)
EFV0062, Work Order 11-34145-
000, Drawing # M13EF05. ID #
PW-1A and PW-2
The inspectors reviewed records for the following welding activities:
SYSTEM WELD IDENTIFICATION WELD TYPE
Reactor Coolant Report # Work Order 339304-000, Tungsten Inert Gas
System ID # W-2 and W-3A, Safety relief Welding (GTAW)
valve drain line valve BBV0085
Reactor Coolant Report # Work Order 339304-005, Tungsten Inert Gas
System ID # W-2B and W-3, Safety relief Welding (GTAW)
valve drain line valve BBV0088
The inspectors verified, by review, that the welding procedure specifications and the
welder had been properly qualified in accordance with ASME Code,Section IX,
requirements. The inspectors also verified, through observation and record review, that
essential variables for the welding process were identified, recorded in the procedure
qualification record, and formed the bases for qualification of the welding procedure
specifications. Specific documents reviewed during this inspection are listed in the
attachment.
These actions constitute completion of the requirements for Section 02.01.
b. Findings
No findings were identified.
.2 Pressurized-Water Reactor Vessel Upper Head Penetration Inspection Activities
(71111.08-02.02)
a. Inspection Scope
During Wolf Creek Refueling Outage 19, a visual examination (VT-2) of the reactor
pressure vessel head was performed. The examination was in accordance with Code
Case N-729-1, Table 1, Item B4.20.
- 14 - Enclosure 2
Also, during the refueling outage, ultrasonic examinations of all seventy-eight control rod
drive mechanism penetration nozzles, and the eddy current examination of the vent line
in the reactor vessel head, was completed. No indications of primary water stress
corrosion cracking were identified. A number of thermal sleeves were found to have
wear extending up to as much as 360 degrees around the thermal sleeve where the
thermal sleeve exits the bottom end of the control rod drive mechanism head adapter
tube. Wear was found in rodded and unrodded penetration locations. The wear was
attributed by the licensee to the thermal sleeve contacting the inside diameter of the
control rod drive mechanism head adapter tube due to a flow-induced impact/whirling
motion of the thermal sleeve. The sleeve-to-adapter contact resulted in wear of material
on the outside diameter of the thermal sleeves.
The licensee informed the inspectors that no immediate remedial action was required.
The inspectors reviewed the licensees evaluation, analysis, and calculations and
concurred with their conclusions. The unrodded thermal sleeves in penetration locations
62 and 63 will need follow-up inspection and/or replacement. From the outer diameter
wear results, the sleeve in location 62 has a predicted life of three inspection cycles, and
the sleeve in location 63 has a predicted life of two inspection cycles. Therefore, the
licensee recommended that these two sleeves be inspected during future refueling
outages for emergent wear.
These actions constitute completion of the requirements for Section 02.02.
b. Findings
No findings were identified.
.3 Boric Acid Corrosion Control Inspection Activities (Pressurized-Water Reactors)
(71111.08-02.03)
a. Inspection Scope
The inspectors evaluated the implementation of the licensees boric acid corrosion
control program for monitoring degradation of those systems that could be adversely
affected by boric acid corrosion. The inspectors participated in containment walkdowns
for identifying locations of boric acid leakage, and reviewed the documentation
associated with the licensees boric acid corrosion control walkdowns as specified in
Procedures STN PE-040D and AI 16F-002. The inspectors also reviewed the visual
records of the components and equipment. The inspectors verified that the visual
inspections emphasized locations where boric acid leaks could cause degradation of
safety-significant components. The inspectors also verified that the engineering
evaluations for those components, where boric acid was identified, gave assurance that
the ASME Code wall thickness limits were properly maintained. The inspectors
confirmed that the corrective actions performed for evidence of boric acid leaks were
consistent with requirements of the ASME Code. Specific documents reviewed during
this inspection are listed in the attachment.
- 15 - Enclosure 2
These actions constitute completion of the requirements for Section 02.03.
b. Findings
.1 Failure to Follow Station Procedures.
Introduction. The inspectors identified a Green non-cited violation of 10 CFR Part 50,
Appendix B, Criterion V, Instructions, Procedures, and Drawings, for failure to follow
procedures to accurately identify, characterize, and resolve boric acid leaks.
Specifically, the licensee failed to recognize discrepancies between boric acid control
procedures and failed to follow boric acid program procedures. Specifically, the licensee
failed to: (1) resolve discrepancies within the boric acid corrosion control program
procedure; (2) resolve discrepancies between the boric acid corrosion control program
procedure and the boric acid leak management procedure; and (3) failed to track and
resolve leakage for locations where health physics had installed drip catch
containments, to review the Health Physics Drip Bag Log as part of the quarterly outside
containment walkdown, and to add component locations to the program. Additionally,
the licensee failed to periodically assess the effectiveness of the program on a refueling
frequency.
Description. The inspectors reviewed station procedures AP 16F-001, Boric Acid
Corrosion Control Program, and AI 16F-002, Boric Acid Leakage Management. In
Procedure AP 16F-001, Attachment A, Section A.1, the least severe leakage where
dry boron residue is present is titled, Non-active Leak, and also classifies leaks into
four categories of severity (Non-Active, Small, Medium, and Large). In Section A.3,
however, the least severe leakage is titled, Inactive Leak. In Procedure AI 16F-002,
Attachment A.1, there are five levels of leakage severity (Non-active, Detectable, Small,
Medium, and Large). This procedure also directs screening/evaluation of boric acid
leakage deficiencies be completed per AI 16F-001, Evaluation of Boric Acid Leakage
(Steps 6.1.1.3, 6.1.2.3). However, the flowchart on Figure 1 references
screening/evaluation per AP 22A-001, Screening, Prioritization, and Pre-Approval,
Revision 15. The inspectors concluded that these procedures provided inconsistent
guidance that affected the licensees ability to properly classify and evaluate boric acid
leaks.
In Procedure AI 16F-002, Steps 5.2.5 and 6.1.5, require the program owner to track and
resolve leakage for locations where health physics had installed drip catch
containments, to review the Health Physics Drip Bag Log as part of the quarterly outside
containment walkdown, and to add component locations to the program. However, the
inspectors noted that consolidation of the health physics log into the Leak Management
program was not regularly completed or documented.
Additionally, steps 6.4.5 and 6.4.6 require the boric acid corrosion control program
owner to periodically assess the effectiveness of the program on a refueling frequency,
including actual performance versus program goals, recommendations for improvement,
summary of inspections/activities performed since last assessment, and a benchmarking
activity once per fuel cycle. However, the inspectors noted that self-assessments were
only completed for the quarterly outside containment walkdowns, and without
- 16 - Enclosure 2
identification of program goals. The inspectors concluded that the licensee was not
performing benchmarking and assessment activities as required by their Boric Acid
Corrosion Control Program.
The inspectors also noted a problem in the frequency of reevaluating past screenings.
Procedure AI 16F-002, Step 6.1.2.3.a. stated that screenings/evaluations for
components with current acid leakage/residue should be updated when the
screening/evaluation is more than 18 months old. However, the inspectors noted that
the station had current acid leakage/residue screenings/evaluations that had not been
updated after 18 months to assess if conditions were still bound by previous evaluations.
The inspectors noted that several condition reports indicated boric acid leakage
locations that had not been adequately identified or evaluated. Condition Report 38972,
initiated on May 9, 2011, indicated boron in the A spent fuel pool cooling pump room
sump. Multiple paths of in-leakage were listed as possible contributing causes to this
accumulation and the work request was closed without resolution of which path(s) were
leaking into the sump. The condition was considered expected and acceptable by
station personnel. Condition Report 60942 reported a large amount of boron build up
around the packing gland of spent fuel pool cleanup pump B, but noted that the same
condition was documented in previous Work Orders 12-356716-000 and Work
Request 12-095525. Condition Report 36024, initiated on March 29, 2011, reported a
leak in the refuel pool. The condition report listed the leak as low significance and not
expected to challenge the function of the refuel pool level limit. The refueling pool was
considered operable but degraded, and the condition report stated that leakage from the
refueling pool had been identified in the past, but the source of leakage was never
identified or evaluated.
The violation was entered into the licensees corrective action program as Condition
Report 65212.
Analysis. The inspectors determined that the failure to recognize discrepancies between
different boric acid control procedures, and the failure to follow boric acid program
procedures was a performance deficiency. The performance deficiency was more than
minor because it affected the Initiating Events Cornerstone attribute of procedure quality
and affected the cornerstone objective to limit the likelihood of those events that upset
plant stability and challenged critical safety functions during shutdown as well as power
operations, and if left uncorrected, the performance deficiency had the potential to lead
to a more significant safety concern. Specifically, failure to resolve discrepancies within
procedures or track and resolve leak locations where health physics had installed drip
catch containments had the potential to mischaracterize leaks or allow leaks to corrode
safety-related systems. Using Inspection Manual Chapter 0609, Appendix A, The
Significance Determination Process for Findings At-Power, the finding was determined
to be of very low safety significance (Green), because the finding was a procedure
quality problem that did not represent a loss of system safety function, and did not
screen as potentially risk significant due to a seismic, flooding, or severe weather
initiating event. The finding had a cross-cutting aspect in the area of human
performance associated with the work practices component because the licensee failed
- 17 - Enclosure 2
to ensure supervisory and management oversight of work activities, including procedure
appropriateness and compliance, such that nuclear safety is supported H.4(c).
Enforcement. Title 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures,
and Drawings, states, in part, activities affecting quality shall be prescribed by
procedures of a type appropriate to the circumstance and accomplished in accordance
with these procedures. Contrary to the above, the licensee failed to prescribe activities
affecting quality by procedures of a type appropriate to the circumstance, and failed to
accomplish activities affecting quality in accordance with procedures. Specifically, the
licensee failed to recognize discrepancies between boric acid control procedures and
failed to follow boric acid program procedures Specifically, prior to February 19, 2013,
the licensee failed to: (1) resolve discrepancies within the boric acid corrosion control
program procedure; (2) resolve discrepancies between the boric acid corrosion control
program procedure and the boric acid leak management procedure; and (3) failed to
track and resolve leakage for locations where health physics had installed drip catch
containments, to review the Health Physics Drip Bag Log as part of the quarterly outside
containment walkdown, and to add component locations to the program. Further, the
licensee had failed to periodically assess the effectiveness of the program on a refueling
frequency. Because this finding was of very low safety significance, it was treated as a
Green non-cited violation in accordance with Section 2.3.2 of the NRC Enforcement
Policy. The violation was entered into the licensees corrective action program as
Condition Report 65212: NCV 05000482/2013003-01, Failure to Follow Station
Procedures.
.2 Failure to Identify Leakage at Refueling Pool Cavity.
Introduction. The inspectors identified Green non-cited violation of 10 CFR Part 50,
Appendix B, Criterion XVI, Corrective Action, for failure to assure that conditions
adverse to quality are promptly identified and corrected. Contrary to the above, the
licensee failed to identify and evaluate a condition adverse to quality in a timely manner.
Specifically, the licensee failed to document the large area of boric acid leakage and
corroded steel plates on the south primary shield wall of the containment refueling pool.
Description. During a boric acid walkdown on February 19, 2013, accompanied by the
licensees program owner, the NRC inspector noted a large area on the backside of the
refueling pool where residue existed around the perimeter of several steel plate
imbedments on two concrete walls and the ceiling that had not been previously identified
by the licensee. The residue had the appearance of a boric acid leak, and one of the
corners of the plates had noticeable corrosion. Procedure AP 16F-001, Boric Acid
Corrosion Control Program, Revision 6B, Step 6.2.1 stated, Sources of boron
seepage/leakage identified by plant personnel per 6.1.1 shall have the following actions
taken as applicable. The large area found on the exterior walls of the refuel cavity,
along with the corroded metal, were reasonable indications that a leak had been
occurring for a considerable amount of time, and should have been noted by station
personnel, as the area was easily accessible and traveled by personnel during refueling
outages. The licensee sent a sample of the residue off site to be analyzed. The results
of the sample identified the residue as boric acid. The licensee concluded that the boric
- 18 - Enclosure 2
acid residue was the result of leakage from the containment refueling pool with migration
through the primary shield wall concrete via construction joints and cracks.
This finding was entered into the licensees corrective action program as Condition
Report 64213.
Analysis. The inspectors determined that the failure to promptly identify and evaluate
the condition adverse to quality was a performance deficiency. The performance
deficiency was more than minor because it affected the Initiating Events Cornerstone
attribute of human performance and affected the cornerstone objective to limit the
likelihood of those events that upset plant stability and challenge critical safety functions
during shutdown as well as power operations and, if left uncorrected, the performance
deficiency would have the potential to lead to a more significant safety concern.
Specifically, failure to implement corrective actions could result in increased leakage and
further degradation of the safety system. The finding has a cross-cutting aspect in the
area of human performance associated with the work practices component because the
licensee failed to define and effectively communicate expectations regarding procedural
compliance and that personnel follow procedures H.4(b).
Enforcement. The inspectors identified a Green non-cited violation of 10 CFR Part 50,
Appendix B, Criterion XVI, Corrective Action, which states, in part, Measures shall be
established to assure that conditions adverse to quality are promptly identified and
corrected. Contrary to the above, the licensee failed to assure that conditions adverse
to quality were promptly identified and corrected. Specifically, prior to February 19,
2013, the licensee failed to document the large area of boric acid leakage and corroded
steel plates on the south primary shield wall of the containment refuel pool. Because
this finding was of very low safety significance, it was treated as a Green noncited
violation in accordance with Section 2.3.2 of the NRC Enforcement Policy. This finding
was entered into the licensees corrective action program as Condition Report 64213:
NCV 05000482/2013003-02, Failure to Identify Leakage at Refuel Pool Cavity.
.3 Steam Generator Tube Inspection Activities (71111.08-02.04)
a. Inspection Scope
The inspectors reviewed the licensees in-situ pressure testing screening criteria for
flawed steam generator tubes to verify that it was in accordance with the EPRI
Guidelines. The inspectors also reviewed the steam generator tube eddy current
examination scope and expansion criteria to determine verify that these meet technical
specification requirements. Also reviewed was the licensees inspection of the
secondary side of the steam generators, and review of the licensees corrective action
taken in response to any observed degradation. The licensee did repairs on select
tubes (e.g., installed plugs or sleeves), and the inspectors observed a portion of these
repairs. The inspector observed the licensees vendor to determine if the equipment was
qualified for detection and/or sizing of the expected types of tube degradation. The
inspectors observed the licensees vendor performing analysis of the steam generator
tubes to determine if proper eddy current testing analysis techniques were applied.
- 19 - Enclosure 2
Wolf Creek is a four-loop plant with Model F steam generators. Each steam generator
includes nominally 5626 tubes made of Alloy 600 thermally treated (A600TI) material.
Wolf Creek had implemented an inspection plan in the past which had exceeded
industry inspection requirements. Prior to Refueling Outage 18, the inspection scope
was 100 percent in two steam generators each outage. The plan was changed to a
sampling inspection in all four steam generators each outage in accordance with the
results of the economic model (Letter SGMP-11-27, Justification of Change to
Inspection Plan for Wolf Creek, dated March 21, 2011).
The primary side inspection scope performed in all four steam generators for the current
outage (Refuel Outage 19) included the following:
- 25 percent bobbin examination of tubes in all four steam generators
- 25 percent hot leg rotating pancake coil tube sheet +3"/-15.21"
- Cold leg peripheral tubes, tube sheet cold +/- 3" 100 percent of peripheral tubes
- +Point examination of all "1-code" indications that are not resolved after history
review
- +Point inspection to bound (all surrounding tubes, at least 1 pitch removed) the
tubes with possible loose part signals during the current inspection
- +Point inspection of possible loose part signals from the previous inspection as
specified in Section 3.5
- 25 percent Row 1 and Row 2 U-bends, mid-range +Point examination
- Dents (structures) >5 volts: Inspect 50 percent in steam generator band C, and
25 percent in steam generators A and D of all previously identified and all new
dents >5 volts in the hot leg (including the U-bends) with the mid-range +Point
probe in all four steam generators
- Dings (free-span) >5 volts: Inspect 25 percent of all previously identified and all
new dings >5 volts in the hot leg (including the U-bends) with the mid-range
+Point probe in all four steam generators. A "new" ding is defined as one for
which there is no prior historical record
- 100 percent bobbin inspection of all prior indications except dents and dings
- +Point examination of a 5 percent sample of bobbin indications that have not
changed since the prior inspection ("H" and "S" codes)
- +Point inspection of the sample of tubes to support the scale profiling effort
- 20 - Enclosure 2
- I00 percent bobbin inspection of tubes identified as potentially having high
residual stress
- 100 percent bobbin inspection of active tubes surrounding previously plugged
tubes
- Visual inspections of all plugs, including factory installed plugs, or their
replacements
- Inspection of potentially deleterious foreign objects (2 tubes)
During the inspection of the hot leg tube sheet expansion zone, a circumferential primary
water stress corrosion crack indication was detected in steam generator B. Due to this
indication, detected at row 17, column 89, tube sheet hot -6.26 inches, the hot leg
rotating pancake coil tube sheet inspection (+3" / -15.21") examination scope was
expanded to 100 percent of tubes in steam generator B with bulge or overexpansion
signals. In addition, the examination scope was confirmed to include at least 20 percent
of tubes in the three other steam generators with bulge or overexpansion signals. No
additional indications were detected. The maximum measured depth of the
circumferential primary water stress corrosion cracking indication in steam generator B
at row 17, column 89, was well below the condition monitoring limit; therefore, the
requirements for condition monitoring were satisfied. The tube was plugged and
because the indication is 6.26 inches inside the tube sheet, there is no concern with
lateral movement if the indication grows to result in tube severance. Because the tube is
unpressurized, there is no pull-out force to cause vertical motion. Therefore, there was
no need to stabilize the tube.
As a result of the eddy current inspection, sixteen tubes were plugged during Refueling
Outage 19. Five tubes in steam generator A, four tubes in steam generator B, two tubes
in steam generator C, and five tubes in steam generator D.
These actions constitute completion of the requirements for Section 02.04.
b. Findings
No findings were identified.
.4 Identification and Resolution of Problems (71111.08-02.05)
a. Inspection Scope
The inspectors reviewed 36 condition reports which dealt with inservice inspection
activities. For the majority of the condition reports, the corrective actions identified for
inservice inspection issues were appropriate. As noted in Section 1R08.3.b.2, the
licensee had missed some opportunities to comply with existing procedures in their
corrective action program. The issue identified in Section 1R08.3.b.2 was a condition
adverse to quality that the licensee failed to identify, therefore the concern of a boric acid
leak was never entered into their corrective action program.
- 21 - Enclosure 2
The inspectors had another observation of the licensees corrective action program,
where the licensee failed to properly evaluate industry generated operating experience.
In January 2012, the licensee received Westinghouse Nuclear Safety Advisory Letter 12-
1 (NSAL-12-1), pertaining to Steam Generator Channel Head Degradation. In January
2012, the licensee wrote Condition Report 00048149, referencing the information in the
Westinghouse Advisory Letter. Condition Report 00048149 stated that the information
is applicable to our steam generators; however, it is not an immediate concern. Wolf
Creek has been performing visual inspections of our steam generator channel heads for
many years. Most recently, the channel heads in all four steam generators were
inspected during RF18 with no anomalies identified. The licensee indicated that the
information would be incorporated as enhancements into work packages and
procedures. On February 22, 2013, during the current Refueling Outage RF19, a visual
inspection of steam generator A hot leg resulted in the licensee identifying a rust-colored
stain at the divider plate to channel head weld, towards the channel head side of the
weld. The stain was identified approximately six inches down from the tube sheet.
Following identification of the potential cladding degradation, ultrasonic testing was
attempted of the area from outside the steam generator primary bowl. The first attempt
was unsuccessful utilizing a straight beam ultrasonic testing probe due to interferences
with the steam generator support beam. Subsequently, a 60-degree L Wave ultrasonic
test probe was utilized to characterize the area. The results from the ultrasonic testing
identified the flaw to be approximately 0.1 inch deep by approximately 2 inches long.
No width could be obtained. The licensee classified the steam generator as degraded
but operable, and planned to perform further evaluation during the next scheduled
refueling outage.
Following the examination, the inspectors questioned the licensees results, conclusions,
and future plans. From this discussion, the inspectors identified that the licensee was
not in compliance with the American Society of Mechanical Engineers (ASME) Boiler
and Pressure Vessel Code. On February 28 and March 11, 2013, conference calls were
held with the NRC, the licensee, and Westinghouse, to discuss the issue of the rusted
area, the inspection techniques used to evaluate the flaw, and the licensees
conclusions. Following the initial conference call, the flaw at the edge of the divider
plate-to-channel head weld in steam generator A was evaluated by Wolf Creek Nuclear
Operating Corporation and Westinghouse, in accordance with Section XI, Paragraph
IWB-3510.1 and Table IWB-351 0-1, of the American Society of Mechanical Engineers
(ASME) Boiler and Pressure Vessel Code (ASME Code). The flaw was characterized as
a planar flaw. The licensee will perform a detailed fracture mechanics and fatigue
growth analysis of the flaw during the next operating cycle, in accordance with
Section XI of the ASME Code. The licensee will re-inspect this area during the next
refueling outage to determine the flaw growth rate.
The inspectors also questioned the licensee about historical documentation supporting
the licensees response that no anomalies had been identified during previous visual
inspections of the steam generators. The licensee performed a historical review of
visual inspections performed on steam generator A bowl that were on digital video discs
and noted that the rust spot was not visible in RF018, but was visible in RF017 and
- 22 - Enclosure 2
RF015. Steam generator A had not been inspected during RF016. Also, the rust spot
was visible in both the RF011 and RF07 video recordings. The RF07 (1994) video is the
earliest video recording of this area. The inspectors concluded that the licensee had
information available for review that should have been evaluated when responding to
Condition Report 00048149. The licensee had not utilized information identified in NRC
Inspection Manual Part 9900, Technical Guidance, such as examinations of records,
inservice testing and inspection programs, maintenance activities, operational event
reviews, operational experience reports, vendor reviews, or inspections, in their
response to Condition Report 00048149.
b. Findings
No findings were identified.
1R11 Licensed Operator Requalification Program and Licensed Operator Performance
(71111.11)
.1 Quarterly Review of Licensed Operator Requalification Program
a. Inspection Scope
On June 11, 2013, the inspectors observed a crew of licensed operators in the plants
simulator during requalification training for steam generator tube rupture methodology
changes. The inspectors assessed the following areas:
- Licensed operator performance
- The ability of the licensee to administer the evaluations
- The modeling and performance of the control room simulator
- The quality of post-scenario critiques
- Follow-up actions taken by the licensee for identified discrepancies
On June 11, 2013, the inspectors observed a crew of licensed operators in the plants
simulator during requalification training for inadvertent safety injection actuation. The
inspectors assessed the following areas:
- Licensed operator performance
- The ability of the licensee to administer the evaluations
- The modeling and performance of the control room simulator
- The quality of post-scenario critiques
- Follow-up actions taken by the licensee for identified discrepancies
These activities constitute completion of two quarterly licensed operator requalification
program samples as defined in Inspection Procedure 71111.11.
b. Findings
No findings were identified.
- 23 - Enclosure 2
.2 Quarterly Observation of Licensed Operator Performance
a. Inspection Scope
On April 29, 2013, the inspectors observed the performance of on-shift licensed
operators in the plants main control room. At the time of the observations, the plant was
in a period of heightened activity due to a unit power reduction in support of emergent
work. The inspectors observed the operators performance of the following activities:
- Primary reactivity changes: control rod manipulations and borations
- Secondary load changes: automatic load set changes
- Swap-over from main feed regulating valves to bypass feed regulating valves
- Swap over of plant electrical loads from unit auxiliary transformer to the start-up
transformer
In addition, the inspectors assessed the operators adherence to plant procedures,
including AP 21-001, Conduct of Operations, and other operations department policies.
These activities constitute completion of one quarterly licensed-operator performance
sample as defined in Inspection Procedure 71111.11.
b. Findings
No findings were identified.
1R12 Maintenance Effectiveness (71111.12)
a. Inspection Scope
The inspectors evaluated degraded performance issues involving the following risk
significant systems:
- Post accident monitoring equipment (Regulatory Guide 1.97) - condition report
67570
- Stator cooling water system - Condition Reports 68393 and 68596
- Watertight pressure doors - Condition Report 65884
- High energy line break doors - Condition Report 66874
The inspectors reviewed events such as where ineffective equipment maintenance has
resulted in valid or invalid automatic actuations of engineered safeguards systems and
independently verified the licensee's actions to address system performance or condition
problems in terms of the following:
- 24 - Enclosure 2
- Implementing appropriate work practices
- Identifying and addressing common cause failures
- Scoping of systems in accordance with 10 CFR 50.65(b)
- Characterizing system reliability issues for performance monitoring
- Charging unavailability for performance monitoring
- Trending key parameters for condition monitoring
- Ensuring proper classification in accordance with 10 CFR 50.65(a)(1) or -(a)(2)
- Verifying appropriate performance criteria for structures, systems, and
components classified as having an adequate demonstration of performance
through preventive maintenance, as described in 10 CFR 50.65(a)(2), or as
requiring the establishment of appropriate and adequate goals and corrective
actions for systems classified as not having adequate performance, as described
The inspectors assessed performance issues with respect to the reliability, availability,
and condition monitoring of the system. In addition, the inspectors verified maintenance
effectiveness issues were entered into the corrective action program with the appropriate
significance characterization. Specific documents reviewed during this inspection are
listed in the attachment.
These activities constitute completion of four quarterly maintenance effectiveness
samples as defined in Inspection Procedure 71111.12-05.
b. Findings
No findings were identified.
1R13 Maintenance Risk Assessments and Emergent Work Control (71111.13)
a. Inspection Scope
The inspectors reviewed licensee personnel's evaluation and management of plant risk
for the maintenance and emergent work activities affecting risk-significant and safety-
related equipment listed below to verify that the appropriate risk assessments were
performed prior to removing equipment for work:
- April 7, 2013, weekly risk assessment 13-202
- April 29, 2013, stator cooling water leak forced outage
- 25 - Enclosure 2
- May 23, 2013, emergency diesel generator A fuel oil transfer pump emergent
work
- June 19, 2013, weekly risk assessment revision for Class 1E air conditioning unit
replacement
The inspectors selected these activities based on potential risk significance relative to
the reactor safety cornerstones. As applicable for each activity, the inspectors verified
that licensee personnel performed risk assessments as required by 10 CFR 50.65(a)(4)
and that the assessments were accurate and complete. When licensee personnel
performed emergent work, the inspectors verified that the licensee personnel promptly
assessed and managed plant risk. The inspectors reviewed the scope of maintenance
work, discussed the results of the assessment with the licensee's probabilistic risk
analyst or shift technical advisor, and verified plant conditions were consistent with the
risk assessment. The inspectors also reviewed the technical specification requirements
and inspected portions of redundant safety systems, when applicable, to verify risk
analysis assumptions were valid and applicable requirements were met. Specific
documents reviewed during this inspection are listed in the attachment.
These activities constitute completion of four maintenance risk assessments
and emergent work control inspection samples as defined in Inspection
Procedure 71111.13-05.
b. Findings
No findings were identified.
1R15 Operability Determinations and Functionality Assessments (71111.15)
a. Inspection Scope
The inspectors reviewed the following assessments:
- April 17, 2013, safety injection pumps A & B run with suction valves closed
- May 29, 2013, turbine-driven auxiliary feedwater steam control positioner failure
and replacement
- June 24, 2013, Class 1E air conditioning unit air flow calculation revision
The inspectors selected these operability and functionality assessments based on the
risk significance of the associated components and systems. The inspectors evaluated
the technical adequacy of the evaluations to ensure technical specification operability
was properly justified and to verify the subject component or system remained available
such that no unrecognized increase in risk occurred. The inspectors compared the
operability and design criteria in the appropriate sections of the technical specifications
and Updated Safety Analysis Report to the licensees evaluations to determine whether
- 26 - Enclosure 2
the components or systems were operable. Where compensatory measures were
required to maintain operability, the inspectors determined whether the measures in
place would function as intended and were properly controlled. Additionally, the
inspectors reviewed a sampling of corrective action documents to verify that the licensee
was identifying and correcting any deficiencies associated with operability evaluations.
Specific documents reviewed during this inspection are listed in the attachment.
These activities constitute completion of three operability evaluations inspection samples
as defined in Inspection Procedure 71111.15-05.
b. Findings
No findings were identified.
1R18 Plant Modifications (71111.18)
.1 Permanent Modifications
a. Inspection Scope
The inspectors reviewed key affected parameters associated with energy needs,
materials, replacement components, timing, heat removal, control signals, equipment
protection from hazards, operations, flow paths, pressure boundary, ventilation
boundary, structural, process medium properties, licensing basis, and failure modes for
the permanent modifications listed below.
- Installation of station blackout diesel generators
- Reactor coolant pump passive thermal shutdown seal modification
- Turbine driven auxiliary feedwater pump governor control modification
The inspectors verified that modification preparation, staging, and implementation did
not impair emergency/abnormal operating procedure actions, key safety functions, or
operator response to loss of key safety functions; post modification testing will maintain
the plant in a safe configuration during testing by verifying that unintended system
interactions will not occur; systems, structures and components performance
characteristics still meet the design basis; the modification design assumptions were
appropriate; the modification test acceptance criteria will be met; and licensee personnel
identified and implemented appropriate corrective actions associated with permanent
plant modifications. Specific documents reviewed during this inspection are listed in the
attachment.
These activities constitute completion of three samples for permanent plant
modifications, as defined in Inspection Procedure 71111.18-05.
b. Findings
No findings were identified.
- 27 - Enclosure 2
1R19 Post-Maintenance Testing (71111.19)
a. Inspection Scope
The inspectors reviewed the following post-maintenance activities to verify that
procedures and test activities were adequate to ensure system operability and functional
capability:
- May 28, 2013, turbine-driven auxiliary feedwater steam control positioner
replacement
- June 21, 2013, Class 1E air conditioning unit compressor replacement
- June 22, 2013, spent fuel pool heat exchanger tube plugging
- June 26, 2013, motor-driven auxiliary feedwater pump A suction pressure
transmitter replacement
- June 26, 2013, motor-driven auxiliary feedwater pump A room cooler leak test
The inspectors selected these activities based upon the structure, system, or
component's ability to affect risk. The inspectors evaluated these activities for the
following (as applicable):
- The effect of testing on the plant had been adequately addressed; testing was
adequate for the maintenance performed
- Acceptance criteria were clear and demonstrated operational readiness; test
instrumentation was appropriate
The inspectors evaluated the activities against the technical specifications, the Updated
Safety Analysis Report, 10 CFR Part 50 requirements, licensee procedures, and various
NRC generic communications to ensure that the test results adequately ensured that the
equipment met the licensing basis and design requirements. In addition, the inspectors
reviewed corrective action documents associated with post-maintenance tests to
determine whether the licensee was identifying problems and entering them in the
corrective action program and that the problems were being corrected commensurate
with their importance to safety. Specific documents reviewed during this inspection are
listed in the attachment.
These activities constitute completion of five post-maintenance testing inspection
samples as defined in Inspection Procedure 71111.19-05.
b. Findings
No findings were identified.
- 28 - Enclosure 2
1R20 Refueling and Other Outage Activities (71111.20)
.1 Refueling Outage
a. Inspection Scope
The inspectors reviewed the outage safety plan and contingency plans for the refueling
outage already in progress at the beginning of this inspection period. Inspection
activities covered in this report were conducted March 1-April 16, 2013, to confirm that
licensee personnel had appropriately considered risk, industry experience, and previous
site-specific problems in developing and implementing a plan that assured maintenance
of defense in depth. During the refueling outage, the inspectors monitored licensee
controls over the outage activities listed below.
- Configuration management, including maintenance of defense in depth, is
commensurate with the outage safety plan for key safety functions and
compliance with the applicable technical specifications when taking equipment
out of service
- Clearance activities, including confirmation that tags were properly hung and
equipment appropriately configured to safely support the work or testing
- Installation and configuration of reactor coolant pressure, level, and temperature
instruments to provide accurate indication, accounting for instrument error
- Status and configuration of electrical systems to ensure that technical
specifications and outage safety-plan requirements were met, and controls over
switchyard activities
- Monitoring of decay heat removal processes, systems, and components
- Verification that outage work was not impacting the ability of the operators to
operate the spent fuel pool cooling system
- Reactor water inventory controls, including flow paths, configurations, and
alternative means for inventory addition, and controls to prevent inventory loss
- Controls over activities that could affect reactivity
- Maintenance of secondary containment as required by the technical
specifications
- Refueling activities, including fuel handling and sipping to detect fuel assembly
leakage
- Startup and ascension to full power operation, tracking of startup prerequisites,
walkdown of the drywell (primary containment) to verify that debris had not been
- 29 - Enclosure 2
left which could block emergency core cooling system suction strainers, and
reactor physics testing
- Licensee identification and resolution of problems related to refueling outage
activities
Specific documents reviewed during this inspection are listed in the attachment.
These activities constitute completion of one refueling outage and other outage
inspection sample as defined in Inspection Procedure 71111.20-05.
b. Findings
No findings were identified.
.2 Forced Outage
a. Inspection Scope
The inspectors reviewed the outage safety plan and contingency plans for the forced
outage to repair the Class 1E air conditioning unit A, conducted May 5-13, 2013, to
confirm that licensee personnel had appropriately considered risk, industry experience,
and previous site-specific problems in developing and implementing a plan that assured
maintenance of defense in depth. During the forced outage, the inspectors observed
portions of the shutdown and cooldown processes and monitored licensee controls over
the outage activities listed below.
- Configuration management, including maintenance of defense in depth, is
commensurate with the outage safety plan for key safety functions and
compliance with the applicable technical specifications when taking equipment
out of service
- Clearance activities, including confirmation that tags were properly hung and
equipment appropriately configured to safely support the work or testing
- Licensee identification and resolution of problems related to forced outage
activities
Specific documents reviewed during this inspection are listed in the attachment.
These activities constitute completion of one refueling outage and other outage
inspection sample as defined in Inspection Procedure 71111.20-05.
b. Findings
No findings were identified.
- 30 - Enclosure 2
1R22 Surveillance Testing (71111.22)
a. Inspection Scope
The inspectors reviewed the Updated Safety Analysis Report, procedure requirements,
and technical specifications to ensure that the surveillance activities listed below
demonstrated that the systems, structures, and/or components tested were capable of
performing their intended safety functions. The inspectors either witnessed or reviewed
test data to verify that the significant surveillance test attributes were adequate to
address the following:
- Preconditioning
- Evaluation of testing impact on the plant
- Acceptance criteria
- Test equipment
- Procedures
- Jumper/lifted lead controls
- Test data
- Testing frequency and method demonstrated technical specification operability
- Test equipment removal
- Restoration of plant systems
- Fulfillment of ASME Code requirements
- Updating of performance indicator data
- Engineering evaluations, root causes, and bases for returning tested systems,
structures, and components not meeting the test acceptance criteria were correct
- Reference setting data
- Annunciators and alarms setpoints
The inspectors also verified that licensee personnel identified and implemented any
needed corrective actions associated with the surveillance testing.
- April 3, 2013, integrated diesel generator and safeguards actuation test train A
- 31 - Enclosure 2
- April 4, 2013, integrated diesel generator and safeguards actuation test train B
- April 24, 2013, manual start, synchronization, and loading of emergency diesel
generator A
- May 15, 2013, turbine-driven auxiliary feedwater pump curve determination
(inservice test)
- June 3, 2013, channel operational test of Tavg, T, and pressurizer pressure
protection set one
- June 5, 2013, turbine-driven auxiliary feedwater system inservice valve test
(inservice test)
- June 5, 2013, turbine-driven auxiliary feedwater pump steam isolation inservice
valve test (inservice test)
Specific documents reviewed during this inspection are listed in the attachment.
These activities constitute completion of seven surveillance testing inspection samples
as defined in Inspection Procedure 71111.22-05.
b. Findings
No findings were identified.
2. RADIATION SAFETY
Cornerstones: Public Radiation Safety and Occupational Radiation Safety
2RS2 Occupational ALARA Planning and Controls (71124.02)
a. Inspection Scope
This area was inspected to assess performance with respect to maintaining occupational
individual and collective radiation exposures as low as is reasonably achievable
(ALARA). The inspectors used the requirements in 10 CFR Part 20, the technical
specifications, and the licensees procedures required by technical specifications as
criteria for determining compliance. During the inspection, the inspectors interviewed
licensee personnel and reviewed the following items:
- Site-specific ALARA procedures and collective exposure history, including the
current 3-year rolling average, site-specific trends in collective exposures, and
source-term measurements
- ALARA work activity evaluations/post-job reviews, exposure estimates, and
exposure mitigation requirements
- 32 - Enclosure 2
- The methodology for estimating work activity exposures, the intended dose
outcome, the accuracy of dose rate and man-hour estimates, and intended
versus actual work activity doses and the reasons for any inconsistencies
- Records detailing the historical trends and current status of tracked plant source
terms and contingency plans for expected changes in the source term due to
changes in plant fuel performance issues or changes in plant primary chemistry
- Radiation worker and radiation protection technician performance during work
activities in radiation areas, airborne radioactivity areas, or high radiation areas
- Audits, self-assessments, and corrective action documents related to ALARA
planning and controls since the last inspection
Specific documents reviewed during this inspection are listed in the attachment.
These activities constitute completion of the one required sample as defined in
Inspection Procedure 71124.02-05.
b. Findings
No findings were identified.
2RS4 Occupational Dose Assessment (71124.04)
a. Inspection Scope
This area was inspected to: (1) determine the accuracy and operability of personal
monitoring equipment; (2) determine the accuracy and effectiveness of the
licensees methods for determining total effective dose equivalent; and (3) ensure
occupational dose is appropriately monitored. The inspectors used the requirements in
10 CFR Part 20, the technical specifications, and the licensees procedures required by
technical specifications as criteria for determining compliance. During the inspection,
the inspectors interviewed licensee personnel, performed walkdowns of various portions
of the plant, and reviewed the following items:
- External dosimetry accreditation, storage, issue, use, and processing of active
and passive dosimeters
- The technical competency and adequacy of the licensees internal dosimetry
program
- Adequacy of the dosimetry program for special dosimetry situations such as
declared pregnant workers, multiple dosimetry placement, and neutron dose
assessment
- Audits, self-assessments, and corrective action documents related to dose
assessment since the last inspection
- 33 - Enclosure 2
Specific documents reviewed during this inspection are listed in the attachment.
These activities constitute completion of the one required sample as defined in
Inspection Procedure 71124.04-05.
b. Findings
No findings were identified.
4. OTHER ACTIVITIES
Cornerstones: Initiating Events, Mitigating Systems, Barrier Integrity, Emergency
Preparedness, Public Radiation Safety, Occupational Radiation Safety, and
Security
4OA1 Performance Indicator Verification (71151)
.1 Data Submission Issue
a. Inspection Scope
The inspectors performed a review of the performance indicator data submitted by the
licensee for the first quarter 2013 performance indicators for any obvious inconsistencies
prior to its public release in accordance with Inspection Manual Chapter 0608,
Performance Indicator Program.
This review was performed as part of the inspectors normal plant status activities and,
as such, did not constitute a separate inspection sample.
b. Findings
No findings were identified.
.2 Reactor Coolant System Specific Activity (BI01)
a. Inspection Scope
The inspectors sampled licensee submittals for the reactor coolant system specific
activity performance indicator for the period from the second quarter 2012 through the
first quarter 2013. To determine the accuracy of the performance indicator data reported
during those periods, the inspectors used definitions and guidance contained in NEI
Document 99-02, Regulatory Assessment Performance Indicator Guideline, Revision 6.
The inspectors reviewed the licensees reactor coolant system chemistry samples,
technical specification requirements, issue reports, event reports, and NRC integrated
inspection reports for the period of April 2012 through March 2013 to validate the
accuracy of the submittals. The inspectors also reviewed the licensees issue report
database to determine if any problems had been identified with the performance
indicator data collected or transmitted for this indicator and none were identified. In
addition to record reviews, the inspectors observed a chemistry technician obtain and
- 34 - Enclosure 2
analyze a reactor coolant system sample. Specific documents reviewed are described
in the attachment to this report.
These activities constitute completion of one reactor coolant system specific activity
sample as defined in Inspection Procedure 71151-05.
b. Findings
No findings were identified.
.3 Reactor Coolant System Leakage (BI02)
a. Inspection Scope
The inspectors sampled licensee submittals for the reactor coolant system leakage
performance indicator for the period from the second quarter 2012 through the first
quarter 2013. To determine the accuracy of the performance indicator data reported
during those periods, the inspectors used definitions and guidance contained in NEI
Document 99-02, Regulatory Assessment Performance Indicator Guideline, Revision 6.
The inspectors reviewed the licensees operator logs, reactor coolant system leakage
tracking data, issue reports, event reports, and NRC integrated inspection reports for the
period of April 2012 through March 2013 to validate the accuracy of the submittals. The
inspectors also reviewed the licensees issue report database to determine if any
problems had been identified with the performance indicator data collected or
transmitted for this indicator and none were identified. Specific documents reviewed are
described in the attachment to this report.
These activities constitute completion of one reactor coolant system leakage sample as
defined in Inspection Procedure 71151-05.
b. Findings
No findings were identified.
4OA2 Problem Identification and Resolution (71152)
.1 Routine Review of Identification and Resolution of Problems
a. Inspection Scope
As part of the various baseline inspection procedures discussed in previous sections of
this report, the inspectors routinely reviewed issues during baseline inspection activities
and plant status reviews to verify that they were being entered into the licensees
corrective action program at an appropriate threshold, that adequate attention was being
given to timely corrective actions, and that adverse trends were identified and
addressed. The inspectors reviewed attributes that included the complete and accurate
identification of the problem; the timely correction, commensurate with the safety
significance; the evaluation and disposition of performance issues, generic implications,
- 35 - Enclosure 2
common causes, contributing factors, root causes, extent of condition reviews, and
previous occurrences reviews; and the classification, prioritization, focus, and timeliness
of corrective actions.
These routine reviews for the identification and resolution of problems did not constitute
any additional inspection samples. Instead, by procedure, they were considered an
integral part of the inspections performed during the quarter and documented in
Section 1 of this report.
b. Findings
No findings were identified.
.2 Daily Corrective Action Program Reviews
a. Inspection Scope
In order to assist with the identification of repetitive equipment failures and specific
human performance issues for follow-up, the inspectors performed a daily screening of
items entered into the licensees corrective action program. The inspectors
accomplished this through review of the stations daily corrective action documents.
The inspectors performed these daily reviews as part of their daily plant status
monitoring activities and, as such, did not constitute any separate inspection samples.
b. Findings
No findings were identified.
.3 Semi-Annual Trend Review
a. Inspection Scope
The inspectors performed a review of the licensees corrective action program and
associated documents to identify trends that could indicate the existence of a more
significant safety issue. The inspectors focused their review on repetitive equipment
issues, but also considered the results of daily corrective action item screening
discussed in Section 4OA2.2, above, licensee trending efforts, and licensee human
performance results. The inspectors nominally considered the 6-month period of
October 2012 through March 2013; although, some examples expanded beyond those
dates where the scope of the trend warranted.
The inspectors also included issues documented outside the normal corrective action
program in major equipment problem lists, repetitive and/or rework maintenance lists,
departmental problem/challenges lists, system health reports, quality assurance
audit/surveillance reports, self-assessment reports, and Maintenance Rule assessments.
The inspectors compared and contrasted their results with the results contained in the
licensees corrective action program trending reports. Corrective actions associated with
- 36 - Enclosure 2
a sample of the issues identified in the licensees trending reports were reviewed for
adequacy.
These activities constitute completion of one single semi-annual trend inspection sample
as defined in Inspection Procedure 71152-05.
b. Findings
No findings were identified.
.4 Selected Issue Follow-up Inspection: Over-torquing Events
a. Inspection Scope
The inspectors recognized a potential trend in over-torquing events at Wolf Creek. The
inspectors observed two events: broken support screws on a Class 1E air conditioning
terminal box as well as over-torquing of the bonnet studs on a safety injection system
check valve. The inspectors reviewed the causes identified and actions taken for each
event. The inspectors also reviewed a previous finding written for over torque of the
essential service water strainer cover to stop leakage without consulting the design
bases of the materials and protective coatings. The inspectors performed a search of
the licensees corrective action database and identified three additional potential over-
torquing events within the last four years. The inspectors presented the events to the
licensee. The licensee wrote Condition Report 65799 to perform a basic trend analysis.
The inspectors reviewed the basic trend analysis.
These activities constitute completion of one in-depth problem identification and
resolution sample as defined in Inspection Procedure 71152-05.
b. Findings
No findings were identified.
.5 Selected Issue Follow-up Inspection: Return to Full Qualification Fuel Cycle Carryover
a. Inspection Scope
During a review of items entered in the licensees corrective action program, the
inspectors recognized a corrective action item documenting the status of all open
operable/functional but degraded/non-conforming conditions that were being
re-evaluated at the end of the refueling outage to determine their suitability for deferral
through fuel cycle 20. All open degraded and non-conforming conditions must be re-
evaluated if they are not corrected during the next reasonable opportunity, such as a
refueling or mid-cycle outage, to ensure that the condition will meet all requirements for
safe operation until the next available opportunity to correct the condition.
These activities constitute completion of one in-depth problem identification and
resolution sample as defined in Inspection Procedure 71152-05.
- 37 - Enclosure 2
b. Findings
No findings were identified.
4OA3 Follow-up of Events and Notices of Enforcement Discretion (71153)
.1 (Closed) Licensee Event Report 05000482/2011-008-00: Post-Fire Safe Shutdown
Latent Design Issue May Cause Essential Service Water System Flow Imbalance
The inspector performed an in-office evaluation of circuit modification and engineering
change package documentation.
On July 20, 2011, during a review of the post fire safe shutdown analysis for
valve EFHV0060, ESW Return from Component Cooling Water Heat Exchanger, the
licensee identified a condition where a fire in the control room could cause the valve to
open and could be damaged such that the valve could not be manually closed. The
direct cause was a latent design deficiency that did not ensure that the valve was
isolated and protected from the potential effects of a control room fire. The licensee
verified that an hourly fire watch was in place in the control room and ensured the fire
watch would remain in place until the issue was resolved.
The licensee generated Engineering Change Package 013898 to modify the control
circuit for the valve. This modification rewired the torque and limit switches to ensure
they are not bypassed by a potential control room hot short (Information Notice 92-18
concern) and installed an isolation/close switch, EFHS0060, to isolate the control room
portion of the circuitry and also to close the valve.
The inspector reviewed the control circuitry modification and engineering change
package, and had discussions with the licensee concerning procedure changes that are
needed as a result of the modification. The inspector verified that the torque and limit
switches would not be bypassed by a hot short in the control room portions of the circuit.
The inspector verified that the isolation/close switch effectively isolated the control room
portions of the circuit and inserted redundant fuses into the control circuit for the valve.
No findings were identified and no violation of NRC requirements occurred. This LER is
closed.
.2 (Closed) Licensee Event Report 2013-003-00: Movement of Irradiated Fuel Progressed
After Non-Conservative Decision Making Resulted in Removal of One Source Range
Monitor from Service
The licensee reported that fuel movement was delayed past the scheduled completion
time due to an equipment problem. Scheduled work on a source range nuclear
instrument was begun while still in the refueling operating mode, when both source
range monitors were required to be operable. The inspectors screened this event using
Inspection Manual Chapter 0612 Appendix B and determined that the performance
deficiencies involved were minor,. Because no fuel movement or other reactivity
- 38 - Enclosure 2
manipulations were in progress during the time this instrument was inoperable. No
additional issues were identified. This LER is closed.
.3 (Closed) Licensee Event Report 2013-005-00: Fatigue Failure of Jacket Water Pressure
Switch Diaphragm Results in Loss of the B Diesel Generator
a. Inspection Scope
The licensee reported that on March 13, 2013, emergency diesel generator B was
rendered inoperable by an equipment failure while emergency diesel generator A was
out of service for planned maintenance during a refueling outage. The licensee declared
a Notice of Unusual Event in accordance with station procedures until emergency diesel
generator B was repaired. The inspectors reviewed the event and the cause evaluation
and determined that this event did involve a violation of regulatory requirements. This
licensee event report is closed.
b. Findings
Diesel Generator Pressure Switch Failed Due to Instrument Line Pressure Oscillations
Introduction. A self-revealing, Green non-cited violation of 10 CFR Part 50. Appendix B,
Criterion XVI, Corrective Action, was identified on March 13, 2013. Specifically, the
licensee repeatedly replaced a jacket water pressure transmitter, but failed to correct
pressure oscillations that caused a fatigue failure of a pressure switch diaphragm, which
rendered emergency diesel generator B inoperable.
Description. On March 13, 2013, the reactor was defueled for a planned refueling
outage and the A emergency diesel generator disassembled for planned maintenance.
At 1:34 a.m. the control room received the B diesel generator trouble alarm. The local
operator found the shutdown relay in the control cabinet had actuated and would not
reset. The engine was declared inoperable and Wolf Creek declared a Notice of
Unusual Event for having two onsite electrical sources unavailable. Instrumentation and
controls technicians troubleshooting the condition determined that the control circuitry
was working properly, but a jacket water pressure switch diaphragm had failed and the
water that leaked was shorting and grounding the associated electrical switch, causing a
false positive signal. This pressure switch was used to indicate that the engine was
running, because the system pressure would be generated by an engine-driven pump.
This false signal rendered the engine inoperable because the resulting logic state
indicated the engine was running with no lube oil pressure, which locked in a protective
engine trip, preventing the engine from starting. The pressure switch was repaired and
the engine was tested and returned to service on March 14, 2013, at 2:21 a.m.,
terminating the Notice of Unusual Event. The licensee wrote Condition Report 65624 to
correct and identify the cause of this condition. This condition only affects the engine
while in standby; if the engine is operating the system would continue to run.
A hardware failure analysis performed on the diaphragm identified that the failure
mechanism was low stress, high cycle fatigue. The pressure switch was nearing the end
- 39 - Enclosure 2
of its specified lifetime; however, there was also a specification for the switch not to
exceed 33,000 pressure cycles to avoid diaphragm failure. The licensee was only
counting the diesel generator stops and starts as a single pressure cycle. However a
review of the machinery history found that a known equipment condition of the jacket
water pressure transmitter hunting had been observed since 2002. This condition was
inducing pressure oscillations in the shared instrument line every one or two seconds
when the engine was running. Furthermore, the magnitude of the oscillations grew as
the pressure transmitter hunting conditions worsened over time. The inspectors
concluded that the licensee focused on correcting the apparent source of the pressure
oscillations, but failed to evaluate the effects of the pressure cycles on components
exposed to the same oscillations. The inspectors noted that the transmitter had been
replaced 10 times between 2002 and 2012. Since the replacements also eventually
exhibited this behavior, the licensee recently determined that is indicative of an
underlying design issue, in that the transmitter model was not being used in the intended
application. Wolf Creek was planning a system modification address and permanently
correct this concern long term, but will be controlled through preventive maintenance in
the interim. Wolf Creek also added preventive maintenance activities to monitor the
replacement diaphragm and other interfacing components.
This issue was entered into the licensees corrective action program as Condition
Report 65624
The inspectors noted that having both emergency diesel generators inoperable at the
same time was permitted by technical specifications at the time of the failure, since the
reactor was defueled. Therefore, no required safety function was lost. The inspectors
also noted that declaration of a Notice of Unusual Event was inconsistent with having no
technical specification requirement to have the function available. Further review noted
that Wolf Creek had not adopted industry standard emergency action level guidance,
which would not have required an event declaration in these circumstances. The
licensee stated that they planned to evaluate adopting the latest guidance.
Analysis. Failure to analyze the effects of pressure oscillations in the emergency diesel
jacket water system on interfacing system components is a performance deficiency. The
performance deficiency is more than minor because it affected the equipment
performance attribute of the Mitigating Systems cornerstone objective to ensure the
availability, reliability, and capability of systems that respond to initiating events to
prevent undesirable consequences. Using Inspection Manual Chapter 0609 Appendix A,
Significance Determination Process for Findings At Power, and determined that the
finding screens as very low safety significance (Green) because the finding does not
meet any criteria outlined in the Exhibit 2, Section A. Specifically, the finding is not a
loss of system safety function and did not exceed its technical specification allowed
outage time of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. The inspectors determined that the finding had a cross-cutting
aspect in the area of problem identification and resolution evaluations because the
licensee failed to ensure that issues that potentially affect nuclear safety are fully
evaluated and addressed in a timely manner. In particular, the licensee repeatedly
replaced the pressure transmitter ten times between 2002 and 2013, including five times
in 2011 and 2012, but failed to evaluate the effect of the pressure oscillations on an
affected component with a limited fatigue life P.1(c).
- 40 - Enclosure 2
Enforcement. Title 10 CFR Part 50. Appendix B, Criterion XVI, Corrective Action,
requires, in part, that Measures shall be established to assure that conditions adverse
to quality are promptly identified and corrected. Contrary to the above, between January
26, 2001, and March 13, 2013, the licensee failed to correct a condition adverse to
quality affecting the jacket water system associated with emergency diesel generator B.
Specifically, despite repeatedly replacing the pressure transmitter that was believed to
be the source of the pressure oscillation, the licensee failed to correct the condition, and
as a result, failed to prevent the subsequent fatigue failure of a pressure switch
diaphragm that rendered the system inoperable. Because the finding is of very low
safety significance and was entered into the licensees corrective action program as
Condition Report 65624, the violation will be treated as a non-cited violation in
accordance with Section 2.3.2.a of the NRC enforcement policy.
NCV 05000482/2013003-03, Diesel Generator Pressure Switch Failed Due to
Instrument Line Pressure Oscillations.
.4 Notice of Unusual Event for a Fire Lasting Greater than 15 Minutes on the Auxiliary
Boiler Roof
a. Inspection Scope
At 2:55 p.m. on April 11, 2013, the fire brigade was called to muster in response to a
confirmed fire in the southeast corner stairway of the turbine building. The inspectors
responded to the control room and to the scene of the fire. The fire was put out, but the
fire re-flashed underneath the insulation. Operations personnel secured the auxiliary
boiler, and the brigade moved onto the turbine building roof where the source of the
stairwell fire was identified as the exhaust stack penetration. Suppression was again
used, but thermal imaging cameras continued to identify hot spots as the fire re-flashed
beneath the stack insulation. Offsite local fire departments responded to the site to
assist with and disassembly and suppression activities until 4:48 p.m., when the fire was
confirmed to be out.
The cause of the fire was believed to an improper repair to the building roof. When the
roof was resealed, the roofers were unable to remove all of the tar and roofing materials
around the penetration, and they insulated over it. After approximately 2 months of
prolonged auxiliary boiler operation during the spring refueling outage, enough heat had
conducted through the stack to ignite the roofing debris. The exhaust stack penetration
has since been repaired. The inspectors screened this event using Inspection Manual
Chapter 0612 Appendix B and determined that the performance deficiencies involved
were not more than minor. All reporting requirements of 10 CFR 50.72 were met. The
inspectors assessed this fire brigade response to satisfy the annual brigade sample in
Section 1R05.
b. Findings
No findings of significance were identified.
- 41 - Enclosure 2
.5 Unplanned Positive Reactivity Transient While Swapping Turbine Operating Modes
a. Inspection Scope
The inspectors reviewed the sequence of events associated with an unplanned power
increase that occurred at 9:35 p.m. on May 2, 2013. Control room operators were
increasing power coming out a forced outage to repair a stator cooling water leak. With
the reactor holding at 79 percent power, operators planned to swap the turbine from full
arc steam admission mode (all four steam control valves throttling equally), the mode
used for turbine startup, and into the partial arc mode (three control valves fully open,
one throttling partially closed) used at full power. During the mode swap, the plant
experienced an unexpected power increase of 11 percent.
The inspectors reviewed procedures for reactivity management and reactivity
manipulations, as well as operator statements. The inspectors reviewed the cause of
the event and corrective actions taken. The inspectors reviewed a recent digital
instrumentation and controls modification to the turbine control system implemented in
the spring 2013 refueling outage.
b. Findings
.1 Failure to Update Station Procedures and Train Operators Regarding the Effects of
Design Changes to the Main Turbine Control System
Introduction. A Green self-revealing non-cited violation of Technical Specification 5.4.1a
was identified for the failure to properly update operating procedures and train operators
on the effects of a recently installed modification. Specifically, procedures were not
adequately revised to provide guidance for operating the new Westinghouse Ovation
digital turbine control system. As a result, operators shifted operating modes at a power
level that caused an unexpected 11 percent power increase due to the combined
characteristics of the steam control valves and the turbine control system. Additionally,
operators received training on shifting control modes at low power, where minor
transients occurred, but were not restricted from performing the swap at high power
levels where the transient could be more significant.
Description. The main turbine controls had been replaced in March 2013 as part of a
planned upgrade during a refueling outage. The controls had satisfactorily passed post
maintenance testing under Temporary Procedure TMP 12-016 during the refueling
outage restart two weeks earlier. During the first plant startup after the modification,
operators initiated the control mode swap below 50 percent reactor power. However, on
May 2, 2013, following an unplanned outage, the turbine control mode swap was
initiated at 76 percent power. As a result, reactor power increased 11 percent power
increase from 76 to 87 percent over a period of 5 minutes.
The following morning the licensee contacted the engineers who had prepared the
modification as well as the vendor (Westinghouse Ovation). The licensee learned that
such a transient was not unexpected under open loop controlling conditions; however,
swapping from full to partial arc mode in that condition is not recommended and should
- 42 - Enclosure 2
be avoided by procedure. Performing the full to partial arc swap in the megawatts
electric or steam pressure control modes will not result in a more than minimal
(1-2 percent) power transient because there is feedback in the circuit to limit changes in
turbine load.
The licensee determined that the mode swap was done in the open loop such that the
controller was programmed with the turbine control valve throttling characteristics as a
substitute for a system response feedback loop. The licensee confirmed earlier testing
that showed the valve characteristics were reasonably accurate below 50 percent power,
but were less accurate at higher powers. Since the turbine controller simultaneously
changed the position of all control valves, three valves were opening while the fourth
valve was to throttle down to compensate for the other three valves. The licensee had
demonstrated that the power transient during mode swap below 50 percent power
stayed within +/-3 percent of the initial power level, and returned to the original power
level. However, starting at 76 percent power, the mode swap resulted in a prolonged
power increase.
The inspectors reviewed the procedures and found that Procedure TMP 12-016
Post Modification Main Turbine Control System Generator Startup and Testing,
Revision 7, used for the post installation testing had steps to take the turbine controller
out of open loop mode before swapping from full to partial arc or vice versa. There was
no caution or warning in the procedure not to perform the swap in open loop mode.
However, there was no step, precaution, limitation, or warning in system operating
Procedure STS AC-001, Main Turbine Valve Testing, Revision 7, to remove the
controller from open loop prior to swapping modes. Control room operators were not
familiar with this potential risk from training either.
The inspectors noted that operator training on the new turbine controller had been
conducted below 50 percent power only. Prior to the modification in the spring 2013
refueling outage, the old turbine control system did not allow open loop control to be
selected. The inspectors determined that this was a new failure mechanism or
vulnerability introduced by the modification and should have been identified in the
planning stages and specifically addressed in the close out process specified in
Section 6.3.4 of AP 05-005, Design, Implementation & Configuration Control of
Modifications, by adding appropriate steps and cautions to procedure SYS AC-001.
This issue was entered into the licensees corrective action program as Condition
Report 68711.
Analysis. Failure to update station operating procedures to provide adequate guidance
for design changes to the turbine control system, and failure to adequately train
operators on those design changes, is a performance deficiency. The performance
deficiency is more than minor because it affected the design control, procedure quality,
and human performance attributes of the Initiating Events cornerstone objective to limit
the likelihood of events that upset plant stability and challenge critical safety functions
during shutdown as well as power operations. Using Inspection Manual Chapter 0609
Appendix A, Checklist 1, Initiating Events Screening Questions, the inspectors
determined that the finding was of very low safety significance (Green) because the
- 43 - Enclosure 2
finding did not result in a reactor trip coincident with the loss of mitigation equipment.
The inspectors determined that this finding had a cross-cutting aspect in the area of
human performance area of work control because the licensee did not appropriately
communicate and coordinate during activities in which interdepartmental coordination
was necessary to assure plant and human performance. Specifically, Wolf Creek did not
communicate and coordinate to ensure that procedure guidance and operator training
adequately conveyed the operational impacts and limitations associated with shifting
turbine control modes at different power levels H.3(b).
Enforcement. Technical Specification 5.4.1a requires that programs specified in the
Appendix A to Regulatory Guide 1.33, Revision 2, be established, implemented, and
maintained. Regulatory Guide 1.33, Appendix A, Section 2.f, includes a general plant
operating procedure for changing load and load following. Contrary to the above, from
April 13 to May 2, 2013, the licensee failed to maintain a general plant operating
procedure for changing load. Specifically, procedure GEN-00-004, Power Operations,
Revision 69, were not updated to provide adequate guidance swapping turbine steam
admission configurations following installation of a new turbine control system. Because
this finding is of very low safety significance and was entered into the licensees
corrective action program as Condition Report 68711, it is being treated as a non-cited
violation in accordance with section 2.3.2.a of the NRC Enforcement Policy:
NCV 05000482/2013003-04, Failure to Update Station Procedures and Train Operators
Regarding the Effects of Implemented Design Changes to the Main Turbine Control
System.
.2 Failure to Properly Manage Reactivity Changes when Swapping Turbine Steam
Admission Modes from Full to Partial Arc
Introduction. Inspectors identified a Green non-cited violation of Technical
Specification 5.4.1.a. for the failure to follow Conduct of Operations and Reactivity
Management procedures.
Description. The inspectors responded to an unplanned reactor transient that occurred
on the night of May 2, 2013. The licensee was increasing power from an unplanned
outage to repair a stator cooling water leak. The unexpected power increase occurred
while swapping the mode of turbine from full to partial arc steam admission while
controlling in the open loop (valve position) mode. This mode swap changes the turbine
control valves that throttle steam to go from all four valves equally throttling to three
valves fully open and one valve throttling such that all four valve slowly reposition over a
period of about 2.5 minutes. The main turbine controls had been replaced in March as
part of a planned upgrade during a refueling outage.
During the first plant startup after the modification, operators initiated the control mode
swap below 50 percent reactor power. However, on May 2, 2013, the mode swap was
initiated at 76 percent power. During the 2.5 minute swap, reactor power increased
11 percent power increase from 76 to 87 percent over a period of about 5 minutes. The
power increase caused Tavg to be below the programmed value for Tref, so the control
rods stepped out until fully withdrawn. The power increase continued, resulting in a
- 44 - Enclosure 2
7 degree Tavg-Tref mismatch as the secondary power demand overcooled the reactor.
Primary pressure lowered by 14 psi, and came within 1 psi of the Departure from
Nucleate Boiling technical specification limit of 2220 psi.
The inspectors reviewed plant parameter graphs during the period of the transient as
well as statements by operators and nuclear engineers in the control room at the time of
the transient. The inspectors concluded that operators had conducted a pre-job brief,
but had failed to discuss the expected plant response in detail, and failed to discuss
contingency actions if the plant response was not as expected. During the transient,
operators discussed taking action to terminate the transient, but were unable to
determine if there was a way to stop the valve swap once it started. Instead, they
attempted to verify that plant parameters did not exceed limits as the transient took its
full course.
The inspectors noted that the pre-job brief and operator response was contrary to the
reactivity control program. Specifically, operators failed to define the expected plant
response and have contingency actions ready in case the plant response was not as
expected. Operation of the main turbine controls was a reactivity manipulation as
defined in Procedure AP 19E-002, Reactivity Management Program, Revision 16.
Following the transient, the inspectors determined that the operators did not adequately
investigate the cause of the unexpected system response before continuing with the
power increase. The inspectors determined that the shift staffing that night did not
include a system expert or vendor representative for the new digital turbine controls to
help investigate the system response. After consulting with the reactor engineers and
the Operations Manager, operators concluded that the turbine controls were behaving as
expected and they had proper control, with the exception of the full to partial arc swap.
The inspectors determined that although the hypothesis was later proven to be correct,
they did not have adequate technical basis to show that equipment that impacts
reactivity was not malfunctioning. No functionality assessment or troubleshooting was
performed, and no technical experts were consulted to verify that the turbine control
response was understood before making the decision to raise power to 100 percent.
The licensee did not gain a full understanding of the event until their discussions with
engineers and the vendor (Westinghouse Ovation) the following morning. Wolf Creek
was informed that such a transient was not unexpected under open loop controlling
conditions, and should have been procedurally prohibited. The licensee was able to
replicate the system response in the plant simulator, both below 50 percent and at
76 percent power. The licensee therefore concluded that the system response was not
anomalous, and that continued operation was appropriate.
AP 19E-002 Reactivity Management Program Section 6.1.1 details the program:
The Reactivity Management Program is the systematic and philosophical direction given
to controlling evolutions with the potential to affect the reactivity and/or integrity of
nuclear fuel. This systematic process ensures that:
- 45 - Enclosure 2
- All deliberate reactivity changes are planned and conducted in a controlled
conservative manner
- Unexpected reactivity changes are minimized
- Conservative actions are taken in response to unexpected reactivity changes
- Reactivity-related modifications, analyses, predictions, and procedures are
correct and effectively implemented
Procedure AP 19E-002 Reactivity Management Program section 5.6 specified that
licensed operator responsibilities as follows:
Licensed Operators are responsible for the implementation of the Reactivity
Management Program. They are responsible for control of reactivity and taking
conservative actions to safeguard the integrity of the reactor fuel. Licensed operators
have the authority to terminate any activity in which the effects on reactivity control are
unknown, unexpected, or non-conservative.
Procedure AP 21-001, Conduct of Operations, also provides guidance on reactivity
management:
- Section 6.1.1.1 states, The greatest responsibility of all licensed operators is to
ensure the reactivity condition of the reactor is monitored and conservatively
controlled at all times.
- Section 6.1.3.1 notes In cases of unplanned reactivity evolutions, licensed
operators must promptly take actions to keep power below 100 percent, stop,
evaluate the plant conditions and take the appropriate conservative action.
Licensed operators shall not hesitate to reduce power, stabilize the plant, or trip
the reactor as necessary to protect the reactor core with concurrence from the
Control Room Supervisor.
Appendix A of this procedure specifies the content of reactivity briefs. The guidance
requires a detailed estimated start and stop point for reactor power as well as an
expected rate of change. Furthermore, contingency actions are to be determined
beforehand if these expectations are not met.
Operator statements and interviews indicate that the anticipated transient was about
1.5 percent, not to exceed 2 percent of rated electric and thermal power. No
contingency actions were specified during the brief. The inspectors determined that the
lack of this contingency planning contributed to the magnitude of the transient and lack
of operator response. For example, by not establishing a limit on the expected plant
response (e.g., require action if power increased above the expected 2 percent rise, or
failed to return to the starting power level) operators were unsure whether action was
needed, and defaulted to the generic Technical Specification limits. As a result, they
failed to act, even when the Tavg-Tref mismatch exceeded operational limits, the
- 46 - Enclosure 2
departure from nucleate boiling pressure limit was closely approached, and control rods
were unable to further compensate. The inspectors noted that later guidance from the
vendor established that operators could have terminated the power excursion at any
time simply by returning turbine steam controller to full arc mode, and thus restoring the
hold previously in place.
The inspectors also concluded that the licensees post-job review of transient lacked the
specific information needed to determine whether the turbine control system had
behaved as expected or had malfunctioned prior to deciding to proceed with power
ascension. In both the pre- and post-job cases, a lack of technical understanding about
the proper workings of the turbine controls was not recognized, and although not
deliberately, operators did proceed in the face of uncertainty. This issue was entered
into the licensees corrective action program as Condition Report 68711.
This issue was entered into the licensees corrective action program as Condition
Report 68711.
Analysis. Failure to provide contingency actions for a greater than anticipated reactor
transient in the pre-job reactivity brief is a performance deficiency. The performance
deficiency is more than minor because it affected the human performance attribute of the
Initiating Events cornerstone objective to limit the likelihood of events that upset plant
stability and challenge critical safety functions during shutdown as well as power
operations. The inspectors evaluated the finding using Inspection Manual Chapter 0609
Appendix A, Checklist 1, Initiating Events Screening Questions, and determined that
the finding was of very low safety significance (Green) because the finding did not result
in a reactor trip coincident with the loss of mitigation equipment. The inspectors
determined that this finding had a cross-cutting aspect in the human performance area
of work practices because the licensee failed to communicate human error prevention
techniques, such as holding pre-job briefings, self and peer checking commensurate
with the risk of the assigned task, such that work activities were performed safely, and
personnel do not proceed in the face of uncertainty or unexpected circumstances.
Specifically, control room operators pre-job reactivity brief was not commensurate with
the risk of the assigned task, and station personnel proceeded to further raise power in
the face of uncertainty about the functionality of the turbine control system H.4(a).
Enforcement. Technical Specification 5.4.1.a requires that programs specified in
Appendix A to Regulatory Guide 1.33, Revision 2, be established, implemented, and
maintained. Regulatory Guide 1.33, Appendix A, Section 1.b includes administrative
procedures covering authorities and responsibilities for safe operation and shutdown.
Contrary to the above on May 2, 2013, did not fully implement the authorities and
responsibilities for safe operation and shutdown. Specifically, operators failed to follow
the Procedure AP 21-001, Conduct of Operations, Revision 61, Appendix Section 1.a.
requirement to establish contingency actions in advance to a planned reactivity
manipulation, in the event that the reactivity addition should exceed the planned amount.
Because this finding is of very low safety significance and was entered into the
licensees corrective action program as Condition Report 68711, it is being treated at a
non-cited violation in accordance with section 2.3.2.a of the NRC Enforcement Policy:
- 47 - Enclosure 2
NCV 05000482/2013003-05, Failure to Properly Manage Reactivity Changes when
Swapping Turbine Steam Admission Modes from Full to Partial Arc.
.6 Unplanned Shutdown due to Non-Functional Class 1E Air Conditioning Unit
On the evening of May 6, 2013, Wolf Creek station operators observed an increasing
trend in the temperature of the train A Class 1E AC and DC switchgear rooms.
Troubleshooting identified that a blockage was present in the thermal expansion valves
that was restricting refrigerant flow. The air conditioning unit itself was declared non-
functional. For systems needed to support the Class 1E AC and DC sources,
Wolf Creek was required to enter the applicable multiple technical specifications. Having
two inverters inoperable was not covered by a specific action statement, so the licensee
appropriately entered Technical Specification 3.0.3, and the unit was shut down to
Mode 5 so that the refrigerant system could be cleaned. Wolf Creek also used this
outage to replace the compressor in this air conditioning unit and chemically clean the
refrigerant system. No findings were identified.
.7 (Open) Notice of Enforcement Discretion (NOED) 13-4-002 for a Non-Functional Class
1E Air Conditioning Unit
On June 17, 2013, an oil sample taken from the train A Class 1E air conditioning unit
was found to have unacceptable levels of aluminum particulate, indicating that internal
parts were degrading and long term reliability was not assured. The unit was declared
non-functional and Wolf Creek again entered Technical Specification 3.0.3 at 11:11 a.m.
Wolf Creek requested a NOED that was granted by the NRC staff at 4:07 p.m. The
inspectors reviewed the documentation, plant status information, the equipment history,
as well as the Inspection Manual Chapter 0410 process. Consistent with NRC policy,
the NRC agreed not to enforce compliance with the specific technical specifications in
this instance, but will further review the cause(s) that created the apparent need for
enforcement discretion to determine whether a violation of NRC requirements existed.
This will be tracked under unresolved item (URI)05000482/2013003-06,
NOED 13-4-002 for a Non-functional Class 1E Air Conditioning Unit.
.8 (Closed) Licensee Event Report 2009-005-01, Loss of Both Diesel Generators with All
Fuel in the Spent Fuel Pool
The inspectors reviewed this LER and determined that the changes to the cause had
already been presented to and inspected by the 95001 inspection team. The results of
this inspection can be found in inspection report 05000482/20130 (ADAMS Accession
Number ML13126A197). LER 2009-005-01 is closed.
4OA5 Other Activities
Falsification of Spent Fuel Pool Area Housekeeping Inspection Records
a. Inspection Scope
The inspectors reviewed procedures and records associated with the conduct and
- 48 - Enclosure 2
completion for housekeeping inspections and foreign material exclusion from important
systems. The inspectors interviewed licensee staff, reviewed inspection records, station
procedures and security card reader logs.
b. Findings
Introduction. The inspectors identified a Severity Level IV violation of 10 CFR 50.9,
Completeness and Accuracy of Information, for the failure to maintain complete and
accurate records required by a license condition. Specifically, the licensee failed to
maintain complete and accurate records of the spent fuel pool area housekeeping
inspections for the period of October through December 2008, required by License
Condition 2.C.5, Fire Protection.
Description. On January 17, 2009, inspectors identified a non-cited violation of Technical
Specification 5.4.1.a, Procedures, for the failure to follow Procedure AP 12-003, Foreign
Material Exclusion, during a walkdown of the spent fuel pool area. During this walkdown,
the inspectors identified numerous untracked tools, equipment and duct tape attached to
various tools such that duct tape was located above and below the fuel pool water level.
In response to the violation, licensee Quality Assurance personnel conducted a review of
the problem that led to the NRC violation. During the review it was identified that the
housekeeping inspection reports had not identified the condition in the spent fuel pool
area. The licensee reviewed the fuel building card reader logs and determined that the
individual assigned to perform the housekeeping inspection in the spent fuel pool area had
not entered the area for a period of 3 months. The licensee determined that the individual
(a supervisor) had reported the information based on feedback received from others who
performed the actual housekeeping observation. Revision 6C of the procedure allowed for
a designee to perform the inspection instead of the management assigned individual. The
ability to designate another individual to perform the inspection was removed when the
procedure was updated in Revision 7 on November 10, 2008, and therefore not allowed
when the December 2008 inspection was completed. In response, the licensee revised
the procedure again to allow for a qualified designee to perform the inspections in
Revision 8, which was issued on September 17, 2009.
NRC inspectors identified that the individual questioned workers who had been in the area
about the condition of the housekeeping before signing off on the inspection with no
issues during the months of October and November 2008. The individual rationalized that
he had met the intent of the inspection, but failed to ask specific questions to ensure that
the inspection met the criteria stated in Attachment B, Building/Area Inspection Checklist.
The inspectors also determined that the individuals had not been designated to perform
the housekeeping inspection, or even told why they were being asked about the condition
of the area. Additionally, the inspectors identified that only one of the people that the
individual had questioned had actually entered the spent fuel pool area in the month of
October 2008. The person that the individual identified as having been questioning to
determine the status of the spent fuel pool area in November 2008, did not access the
spent fuel pool area that month. On December 4, 2008, the individual documented the
completion of an inspection with no issues in the Housekeeping Inspection Card without
- 49 - Enclosure 2
questioning any individuals or entering the area. The individual stated he knew work had
not been performed in the area since the previous report on November 25, 2008, so he
assumed that the condition remained unchanged.
The NRC determined that from October through December 2008, a licensee employee
failed to perform inspections of the spent fuel pool area in accordance with Procedure
AP 12-001, Housekeeping Control, and willfully documented false information on the
Housekeeping Inspection Cards.
The falsified inspection report was related to a housekeeping inspection procedure that
implements housekeeping inspections required by the fire protection program. The Wolf
Creek Updated Final Safety Analysis Report describes the fire protection program in
Section 9.5-1, and commits to meeting Regulatory Guide 1.39, Housekeeping
Requirements for Water-Cooled Nuclear Power Plants, Revision 2, Appendix 3A,
Conformance to NRC Regulatory Guides. Regulatory Guide 1.39, Revision 2, endorses
ANSI Standard N45.2.3-1973, Housekeeping During the Construction Phase of Nuclear
Power Plants, as a method of complying with fire protection program housekeeping
requirements. The ANSI Standard requires that periodic inspection and examination of
the work areas shall be performed at scheduled intervals to assure adequacy of
cleanliness and housekeeping practices, and that copies of inspection and examination
records shall be prepared and placed with other project records.
Analysis. The failure to maintain records required by License Condition that are complete
and accurate in all material respects in accordance with 10 CFR 50.9 was a violation.
Because the violation is associated with willfulness and impacted the regulatory process it
was evaluated under the traditional enforcement process as set forth in the NRC
Enforcement Policy. Since this violation was the result of a willful action, the NRC
considers the violation to be more than minor, and therefore, the NRC has classified the
violation at Severity Level IV, in accordance with the NRC Enforcement Policy (Section
4OA5).
Enforcement. Title 10 CFR 50.9 requires, in part, that information required by statute,
orders, or license conditions to be maintained by the licensee shall be complete and
accurate in all material respects.
Wolf Creek License Condition 2.C.5, Fire Protection, requires that the licensee shall
implement and maintain in effect all provisions of the approved fire protection program as
described in the SNUPPS Final Safety Analysis Report. The Wolf Creek Updated Final
Safety Analysis Report describes the fire protection program in Section 9.5-1, and
commits to meeting Regulatory Guide 1.39, Housekeeping Requirements for Water-
Cooled Nuclear Power Plants, Revision 2, in Appendix 3A, Conformance to NRC
Regulatory Guides, and Appendix A, Table 9.5A-1. Regulatory Guide 1.39, Revision 2,
endorses ANSI Standard N45.2.3-1973, Housekeeping During the Construction Phase of
Nuclear Power Plants.
ANSI Standard N45.2.3-1973, Section 3.5 states, in part, that periodic inspection and
examination of the work areas shall be performed at scheduled intervals to assure
- 50 - Enclosure 2
adequacy of cleanliness and housekeeping practices. Section 4 states, in part, that
copies of inspection and examination records shall be prepared and placed with other
project records.
Procedure AP 12-001, Housekeeping Control, Revision 6C, dated May 5, 2006,
Section 6.1.8 required that Assigned personnel shall walk down their areas monthly
(3) These Individuals or Designees shall walk down their assigned areas monthly.
Procedure AP 12-001, Housekeeping Control, Revision 7, dated November 10, 2008,
Section 6.1.8, required that Assigned personnel shall walk down their areas monthly
[i.e., allowed use of designees was removed].
Contrary to the above, between October and December 2008, the licensee failed to
maintain records required by License Condition 2.C.5 that were complete and accurate in
all material respects. Specifically, the Housekeeping Inspection Card for the spent fuel
pool area indicated that the inspection had been completed when security access logs
indicate that the individual that completed the record had not entered the area. The NRC
investigation determined that the assigned individual did not walk down the assigned area,
and did not assign a designee to do so.
This is a violation of 10 CFR 50.9. A notice of violation is attached. NOV 05000482/2013-
08 Failure to Maintain Complete and Accurate Housekeeping Records. (EA-13-084)
4OA6 Meetings, Including Exit
Exit Meeting Summary
On May 23, 2013, the inspectors presented the results of the radiation safety inspections to
Ms. A. Stull, Vice President and Chief Administrative Officer, and other members of the licensee
staff. The licensee acknowledged the issues presented. The inspectors asked the licensee
whether any materials examined during the inspection should be considered proprietary. No
proprietary information was identified.
On June 12, 2013, the inspector debriefed the results of the review of LER 2011-008-00 to
Mr. R. Hobby, Licensing, and Mr. D. Garbee, Acting Fire Protection Supervisor. The licensee
acknowledged the issues presented. No proprietary information was reviewed.
On July 11, 2013, the resident inspectors presented the inspection results to Mr. J. Broschak,
and other members of the licensee staff. The licensee acknowledged the issues presented.
The inspector asked the licensee whether any materials examined during the inspection should
be considered proprietary. No proprietary information was identified.
On August 7, 2013, the resident inspectors conducted a supplemental exit with Mr. R.Smith to
revise the characterization of two findings. The licensee acknowledged the issues presented.
The inspector asked the licensee whether any materials examined during the inspection should
be considered proprietary. No proprietary information was identified.
- 51 - Enclosure 2
SUPPLEMENTAL INFORMATION
KEY POINTS OF CONTACT
Licensee Personnel
A. Camp, Plant Manager
A. Stull, Vice President and Chief Administrative Officer
B. Fox, Contractor, Fire Protection Engineer
D. Dees, Operations Support Superintendant
D. Grove, Maintenance Superintendant
E. Peterson, Ombudsman
G. Pendergrass, Manager Station Recovery
J. Broschak, Engineering Vice President
J. Kobyra, Manager Design Engineering
J. Schepers, Supervisor, Radiation Protection
J. Yunk, Manager Performance Improvement and Corrective Actions
K. Davis, Welder, Mechanical Maintenance
L. Aiken, Master Health Physics Technician, Radiation Protection
L. Lane, Operations Superintendant
L. Ratzlaff, Manager Maintenance
L. Upson, Manager Integrated Plant Scheduling
M. Church, Master Welder, Mechanical Maintenance
M. Skiles, Supervisor, Radiation Protection
M. Sunseri, President and Chief Executive Officer
M. Westman, Manager Regulatory Affairs
P. Bedgood, Manager Radiation Protection
P. Herrman, Manager Support Engineering
R. Clemens, Strategic Projects Vice President
R. Flannigan, Manager Nuclear Engineering
R. Hobby, Specialist, Licensing
R. Rumas, Manager Quality
R. Smith, Site Vice President and Chief Nuclear Operations Officer
S. Henry, Manager Operations
T. Baban, Manager System Engineering
T. Damashek, Operations Training Superintendent
T. East, Superintendent of Emergency Planning
T. Patten, Master Health Physics Technician, Radiation Protection
T. Slenker, Operations Corrective Action Program Coordinator
W. Muilenberg, Supervisor Licensing
A1-1 Attachment 1
LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED
Opened
05000482/2013003- Notice of Enforcement Discretion (NOED) 13-4-002 for a Non-
07 Functional Class 1E Air Conditioning Unit (Section 4OA3.7)
Closed
05000482/2011- Post-Fire Safe Shutdown Latent Design Issue May Cause
LER
008-00 Essential Service Water System Flow Imbalance (Section 4OA3.1)
05000482/2009- Loss of Both Diesel Generators with All Fuel in the Spent Fuel
LER
005-01 Pool (Section 4OA3.8)
Movement of Irradiated Fuel Progressed After Non-Conservative
05000482/2013-
LER Decision Making Resulted in Removal of One Source Range
003-00
Monitor from Service (Section 4OA3.2)
05000482/2013- Fatigue Failure of Jacket Water Pressure Switch Diaphragm
LER
005-00 Results in Loss of the B Diesel Generator (Section 4OA3.3)
Opened and Closed
05000482/2013003-
NCV Failure to Follow Station Procedures (Section 1R08.3.b.1)
01
05000482/2013003- Failure to Identify Leakage at Refuel Pool Cavity (Section
02 1R08.3.b.2)
05000482/2013003- Diesel Generator Pressure Switch Failed Due to Instrument Line
03 Pressure Oscillations (Section 4OA3.3)
Failure to Update Station Procedures and Train Operators
05000482/2013003-
NCV Regarding the Effects of Implemented Design Changes to the
04
Main Turbine Control System (Section 4OA3.5.b.1)
Failure to Properly Manage Reactivity Changes when Swapping
05000482/2013003-
NCV Turbine Steam Admission Modes from Full to Partial Arc (Section
05
4OA3.5.b.2)
05000482/2013003- Failure to Maintain Complete and Accurate Housekeeping Records
06 (Section 4OA5)
A1-2
LIST OF DOCUMENTS REVIEWED
Section 1R01: Adverse Weather Protection
PROCEDURES
NUMBER TITLE REVISION
OFN AF-025 Unit Limitations 39
ALR 00-114D OA/OPC Trouble 5
ALR 00-134E Main Transformer Trouble 11
Section 1R04: Equipment Alignment
PROCEDURES
NUMBER TITLE REVISION
SYS KJ-125 EDG Starting Air Compressor Operation 15
CKL EF-120 Essential Service Water Valve, Breaker and Switch Lineup 46
CKL AL-120 Auxiliary Feedwater Normal Lineup 40A
DRAWINGS
NUMBER TITLE REVISION
M-12KJ05 Piping and Instrumentation Diagram, Standby Diesel 16
Generator B Intake Exhaust, F.O. & Start Air Sys.
M-11EF01 System Flow Diagram Essential Service Water 09
M-12AL01 Piping and Instrumentation Diagram Auxiliary Feedwater 23
System
CONDITION REPORTS
54654 62411 62413
WORK ORDERS
07-292792-016 07-292792-017 07-292792-021 07-292792-022 07-292792-039
07-292792-040
A1-3
Section 1R05: Fire Protection
PROCEDURES
NUMBER TITLE REVISION
AP 10-100 Fire Protection Program 17
AP 10-106 Fire Preplans 13
DRAWINGS
NUMBER TITLE REVISION
E-1F9905 Fire hazard Analysis 4
Section 1R08: Inservice Inspection
CONDITION REPORTS
00034867 00036937 00041751 00051596
00035428 00036938 00043489 00053786
00035429 00037276 00043490 00054549
00035665 00037334 00044754 00056232
00035940 00037386 00044963 00056535
00036024 0037803 00044966 00057429
00036438 00038141 00047212 00057463
00036443 00038972 00048307 00058234
00036876 00041750 00048368 00058734
DRAWINGS
NUMBER TITLE REVISION
M-1G062 Equipment Location Turbine Building Partial Plan 2
El. 2015-4
PROCEDURES
NUMBER TITLE REVISION
AI 16F-001 Evaluation Of Boric Acid Leakage 7
AI 16F-002 Boric Acid Leakage Management 7
AP 16F-001 Boric Acid Corrosion Control Program 6B
AP 22A-001 Screening, Prioritizing, and Pre-Approval 15
AP 29A-003 Steam Generator Management 15
AP 29A-004 American Society Mechanical of Engineers (ASME) 8
Section Xi System Pressure Testing
A1-4
PROCEDURES
NUMBER TITLE REVISION
I-ENG-023 Steam Generator Data Analysis Guidelines 13
PDI ISI 254 SE NB Remote lnservice Examination of Reactor Vessel 2
Nozzle to Safe End, Nozzle to Pipe and Safe End to
Pipe Welds Using the Nozzle Scanner
PDI UT 8 PDI Generic Procedure for Ultrasonic Examination of F
Weld Overlaid Similar and Dissimilar Metal Welds
PDI-UT-1 Generic Procedure for Ultrasonic Examination of E
Ferritic Pipe Welds
QCP-20-502 Magnetic Particle Examination AC/DC Yoke and AC 8B
Coil Techniques
QCP-20-520 Pressure Test Examination 9
STN PE-040D RCS Pressure Boundary Integrity Walkdown 3
STN PE-040G Transient Event Walkdown 4
STN PE-370 Foreign Object Search and Retrieval and Secondary 12
Side Inspections
STS PE-022 Steam Generator Tube Inspection 19
STS PE-040E RPV Head Visual Inspection 3
UT 2 Ultrasonic Examination of Vessel Welds and Adjacent 28
Base Metal
UT 92 Ultrasonic Examination of Overlaid Austinetic Piping 6
Welds
UT-94 Lambert, McGill, and Thomas Nondestructive 7
Examination Procedure - Ultrasonic Examination of
Ferritic Pipe Welds
WDI STD 101 RHVI Vent Tube J-Weld Eddy Current Examination 10
WDI STD 1040 Procedure for Ultrasonic Examination of Reactor 9
Vessel Head Penetrations
WDI STD 1041 Reactor Vessel Head Penetrations Ultrasonic 8
Examination Analysis
WDI STD 114 RHVI Vent Tube ID and CS Wastage Eddy Current 12
Examination
WDI STD 146 ET Examination of Reactor Vessel Pipe Welds Inside 11
Surface
A1-5
MISCELLANEOUS
NUMBER TITLE REVISION / DATE
09-00178 Wolf Creek Generating Station - Request For March 27, 2009
Additional Information Re: Relief Request 13R-06,
Alternative To The Examination Requirements Of
ASME Section XI For Class 1 Piping Welds
Examined From The Inside Of The Reactor Vessel
ASS03 Performance Improvement Learning Oversight and September 20, 2006
Trending System Assessment/Audit Detail Report -
BACCP Self-Assessment
Code Case Alternative Examination Requirements for PWR March 28, 2006
N-729-1 Reactor Vessel Upper Heads With Nozzles Having
Pressure-Retaining Partial-Penetration Welds
Section XI, Division 1
ES1301910 Boric Acid Training for WCNOC Supervision 001
ESH - 102 STARS Plants Alloy 600 Program Review September 5, 2006
ET 06-001 0 Docket 50-482: Inservice Inspection Program Plan March 2, 2006
for the Third Ten-Year Interval and 10 CFR 50.55a
Requests 13R-01, 13R-02,and 13R-04
ET 06-0021 Docket No. 50-482: 10 CFR 50.55a Request 13R- May 9, 2006
05, Installation and Examination of Full Structural
Weld Overlays for Repairing/Mitigating Pressurizer
Nozzle-to-Safe End Dissimilar Metal Welds and
Adjacent Safe End-to-Piping Stainless Steel Welds
ET 06-0031 Docket 50-482: Wolf Creek Nuclear Operating August 4, 2006
Corporation's Response to Request for Additional
Information Regarding I 0 CFR 50.55a Request l3R-
05 and Submittal of Revision 1 to 10 CFR 50.55a
Request 13R-05
ET 060042 Docket 50-482: Wolf Creek Nuclear Operating September 27, 2006
Corporation's Response to the September 20, 2006
NRC Request for Additional Information Regarding
10 CFR 50.55a Request 13R-05
ET 06-0043 Docket 50-482: Wolf Creek Nuclear Operating October 5, 2006
Corporation's Response to NRC Request for
Additional Information Regarding 10 CFR 50.55a
Request 13R-01
ET 06-0044 Docket 50-482: Wolf Creek Nuclear Operating October 2, 2006
Corporations Revised Commitment Regarding 10
CFR 50.55a Request 13R-05
A1-6
MISCELLANEOUS
NUMBER TITLE REVISION / DATE
ET 06-0058 Docket No. 50-482: Wolf Creek Nuclear Operating December 20,2006
Corporation's Response to the Second NRC
Request for Additional Information Regarding 10
CFR 50.55a Request 13R-01
ET 08-0044 Docket No. 50-482: 10 CFR 50.55a Request 13R- September 16, 2008
06, Alternative to the Examination Requirements of
ASME Section XI for Class 1 Piping Welds
Examined from the Inside of the Reactor Vessel
ET 09-0014 Docket No. 50-482: Wolf Creek Nuclear Operating April 23, 2009
Corporation's Response to Request for Additional
Information Regarding 10 CFR 50.55a Request
ET 12-0010 Docket 50-482: 10 CFR 50.55a Request Number July 2, 2012
13R-07, Relief from ASME Code Case N-729-1
Requirements for Examination of Reactor Vessel
Head Penetration Welds
Letter from Wolf Creek Generating Station -Request For Relief January 4, 2013
Matthew W. No. 13R-07 For The Third 10-Year Inservice
Sunseri Inspection Program Interval (TAC No. ME9078)
Letter from Wolf Creek Generating Station - Third 10-Year February 21, 2007
Rick A. Muench Interval Inservice Inspection Program Relief
Request I3R-01 (TAC No. MD0297
Letter from Wolf Creek Generating Station - Authorization Of July 19, 2007
Rick A. Muench Relief Request I3r-05, Alternatives To Structural
Weld Overlay Requirements (TAC No. MD1813)
Letter from Wolf Creek Generating Station -Relief Request July 23, 2009
Rick A. Muench 13R-06, Alternative To The Examination
Requirements Of ASME Code,Section XI For Class
1 Piping Welds Examined From The Inside Of The
Reactor Vessel (TAC No. MD9658)
Letter from NCR Wolf Creek Generating Station -Issuance Of November 19, 2012
to Matthew W. Amendment RE: Adoption of TSTF-510, Revision 2,
Sunseri "Revision To Steam Generator Program Inspection
Frequencies And Tube Sample Selection," Using
The Consolidated Line Item Improvement Process
Letter from NCR Wolf Creek Generating Station -Issuance Of December 11, 2012
to Matthew W. Amendment Re: Steam Generator Tube Permanent
Sunseri Alternate Repair Criteria (TAC No. ME8350)
A1-7
MISCELLANEOUS
NUMBER TITLE REVISION / DATE
ME9078 Request For Additional Information Request I3R-07 September 4, 2012
Examination Of Reactor Vessel Head Penetration
Welds Wolf Creek Generating Station Unit 1 Wolf
Creek Nuclear Operating Corporation Docket
Number 50-482
October 15, 2012 Docket 50-482: Response to Request for Additional October 15, 2012
Information Regarding 10 CFR 50.55a Request
Number 13R-07, "Relief from ASME Code Case N-
729-1 Requirements for Examination of Reactor
Vessel Head Penetration Welds"
SA-2012-0023 ISI Program Self Assessment March 8, 2012
SEL 2010-163 Self-Assessment Report SEL 2010-163 March 25, 2010
Steam Generator Health Optimization
SG-CDME-12-2 Wolf Creek Steam Generator Secondary Side October 2012
Condition Monitoring and Operational Assessment
for Fuel Cycle 19 and Refueling Outage 19
SG-SGDA-11-1 Wolf Creek RF18 Condition Monitoring Assessment January 2012
and Operational Assessment
Section 1R11: Licensed Operator Requalification Program
PROCEDURES
NUMBER TITLE REVISION
AP 21-001 Conduct of Operations 60
GEN 00-004 Power Operations 69
GEN 00-005 Minimum Load to Hot Standby 75
OP3003501 Steam Generator Tube Rupture Response Methodology 00
Change Lab
LR5002023 Inadvertent Safety Injection Lab 004
Section 1R12: Maintenance Effectiveness
PROCEDURES
NUMBER TITLE REVISION
AP 23M-001 WCGS maintenance Rule Program 9
A1-8
CONDITION REPORTS
68393 70858 70854 66874 ACE 51622
66244 66499 66875
WORK ORDERS
12-353867-000 13-099778 13-366342-000 13-366342-001
MISCELLANEOUS
NUMBER TITLE REVISION /
DATE
Maintenance Rule Database
Section 1R13: Maintenance Risk Assessment and Emergent Work Controls
PROCEDURES
NUMBER TITLE REVISION /
DATE
AP 22C-003 Online Nuclear Safety and Generation Risk Assessment 19
APF 22C-003-01 On-Line Nuclear Safety and Generation Risk Assessment May 20, 2013
Week (2013) 208 (as revised)
MISCELLANEOUS
NUMBER TITLE REVISION
Work Week 2013-205 Major Activities (4/29-5/05) 0
Work Week 13-202 Risk Assessment 0
Section 1R15: Operability Evaluations
PROCEDURES
NUMBER TITLE REVISION
AP 06-002 Radiological Emergency Response Plant (RERP) - 13
Emergency Action Level -1 Radioactive Effluent Release
AP 26C-004 Operability Determination and Functionality Assessment 26
AI 22C-016 Unit Condition and Operational Residual Risk 0A
AI 22C-010 Operations Work Controls 15
STS AL-103 TDAFW Pump Inservice Test 57
A1-9
CONDITION REPORTS
00062146 00064397 00067888 66396 66398
00057299 00063495
MISCELLANEOUS
NUMBER TITLE REVISION /
DATE
13-364867-001 Engineering Disposition: Justification of AFW Pump PAL02 March 8,
Stuffing Box Extension Through-Wall 2013
13-365489-001 Engineering Disposition: PEJ01A Diffuser Volute Vane 2
Damage
13-366890-000 & Engineering Disposition: PEM01A/B Operated With No 0
13-366291-000 Suction Sources
Interim Operation Assessment: PEM01A (SI pump A) April 18,
2013
Interim Operation Assessment: PEM01B (SI pump B) April 8, 2013
GK-13-004 Operability Evaluation: SGK05A Class 1E AC Unit A 0
Engineering Disposition: ESC - PEM01A/B Operated With 0
No Suction Sources
Section 1R18: Plant Modifications
TEMPORARY MODIFICATION ORDER
NUMBER TITLE REVISION
13-003-KE Removal of Fuel Transfer System Hold Down Assembly 0
DOCUMENTS
NUMBER TITLE REVISION /
DATE
DCP 14263 SBO Diesel Generator Project- Missile Barrier Erection and 4
Equipment Installation Work Outside Protected Area
DCP 12958 Turbine Driven Auxiliary Feedwater Pump Governor Control 9
Modification
A1-10
DOCUMENTS
NUMBER TITLE REVISION /
DATE
WCAP-17100-P PRA Model for the Westinghouse Shutdown Seal, 0
Supplement 1
WCAP-17541-P Implementation Guide for Westinghouse Reactor Coolant 0
Pump SHIELD Passive Thermal Seal
AI 21-017 Timed Fire Protection Actions Validation 4
MPM M712Q-01 Reactor Coolant Pump Seal Removal/Installation 23
XX-E-013-002 Post-fire Safe Shutdown (PFSSD) Analysis 19
WCOP-24 Operations EMG/OFN Setpoints 11
OFN BB-005 RCP Malfunctions 21
RCP Vendor Technical Manual 6
MGE TL-001 Wiring Termination and Lug/Connector Installation 19
AP 16E-002 Post maintenance Testing Development 13
2013-432 Breach Permit April 15,
2013
OE NB-13-002 Operability Evaluation NB001, NB002 0
QR-03117685-1 Qualification Report for Current Transformer and Modified 2
Bus Installation for General Electric Type M26 Switchgear
Technical Review of NB-13-02 0
AP 10-104 Breach Authorization 27
A1-11
DRAWINGS
NUMBER TITLE REVISION
WIP-E-13NB03- Lower Medium Voltage Sys. Class 1E 4.16KV Three Line 1-6
005-A-1 Meter and Relay Diagram
WIP-E- Schematic Diagram Class 1E Bus NB)! Feeder BRKR. 0
13KUO1A-000- 152NB0114
A-1
WIP-E-009- MetalClad Switchgear Connection Diagram 1
00132-W08-A-1
WIP-E-009- Electrical Diagram: Power Control Circuits 1
00024-W11-A-1
M-712-00206 Shutdown Seal Model 93A-1 & 100 RCP Shutdown Seal 1
Assembly Components
M-712-00207 No. 1 Seal Assembly Kit 8 inch 1
M-712-00056 General Assembly 93A-1 R.C. Pump 16
M-712-00057 General Assembly 93A-1 R.C. Pump 11
M-712-00058 General Assembly 93A-1 R.C. Pump 9
M-712-00059 General Assembly 93A-1 R.C. Pump 15
Work in Progress Drawings Associated with DCP 14117
Work in Progress Drawings Associated with DCP 14261
Work in Progress Drawings Associated with DCP 14262
Work in Progress Drawings Associated with DCP 14263
A1-12
WORK ORDERS
12-350886-025 12-350886-026 12-350886-027 12-350886-028 12-350886-029
12-350886-030 12-350886-031 12-350886-032 12-354257-003 12-354257-063
12-361102-016 12-354257-006 12-354257-126 12-354257-128 12-354257-129
CONDITION REPORTS
66117 65321 63243 66592 66698
MISCELLANEOUS
NUMBER TITLE REVISION
USAR 15.7.4 Fuel Handling Accidents 21
Section 1R19: Post-Maintenance Testing
PROCEDURES
NUMBER TITLE REVISION
MPE BA 014 Battery Impedance Test 4A
MPE E050Q-05 Battery Equalizing Procedure 13A
STS MT-019 125VDC Class 1E Quarterly Battery Inspection 21
STS MT-020 125 Volt DC Battery Inspection/Charger Operational Test 25B
STS MT-021 Service Test for 125Vdc Class 1E Batteries 16A
STS EF-100B ESW Pump B In-service Test and Discharge Check Valve 40
In-service Test
SYS GK-123 Control Building A/C Units Startup and Shutdown 21
STS IC-565 Channel Calibration Auxiliary Feedwater Pump Suction 5A
Pressure Indication for Remote Shutdown Pressure
CNT-MM-700 Fabrication and Installation of Tubing, Tubing Supports, 5
Instrument Supports and Instrument Installation
MPE GK-003 Control Room and Class 1E A/C Units Preventive 4
Maintenance Activity
MPE GK-004 GK Unit Preparation for Work 4
A1-13
CONDITION REPORTS
70420 ACE 19528 67888 66398 66396
57299
WORK ORDERS
11-340517-002 11-341224-001 09-321171-001 11-341337-002 11-342032-004
08-309413-041 09-342741-002 11-341336-003 11-345398-002 09-317266-001
11-343552-002 11-343567-001 12-353040-003 11-345397-002 11-337095-005
11-343332-000 11-343334-000 12-352686-000 13-373150-004 10-333747-000
13-373150-002 12-352686-000 13-373153-001 13-373153-009 12-361695-005
13-373153-008 13-373153-004 13-373153-000
MISCELLANEOUS
NUMBER TITLE REVISION /
DATE
Balance of Plant Eddy Current Inspection Report EEC01A June 20,
Spent Fuel Pool Cooler 2013
M-071-0016-05 Vendor Manual: Cooling the Spent Fuel Pool 3
M-071-0015-03 Vendor Manual: Exchange Surface Requirement Based on May 12, 1977
Case A Conditions
V5011748 SGKOSA Oil Analysis Certificate, Herguth Laboratories June 26,
2013
V5011749 SGKOSA Oil Analysis Certificate, Herguth Laboratories June 26,
2013
Section 1R20: Refueling and Other Outage Activities
PROCEDURES
NUMBER TITLE REVISION
GEN 00-003 Hot Standby to Minimum Load 87A
RXE 01-002 Reload Low Power Physics Testing 24
CONDITION REPORTS
00064552 00063645
A1-14
MISCELLANEOUS
NUMBER TITLE REVISION
Reactivity Maneuver Plan, Cycle 20 Initial Startup 0
Section 1R22: Surveillance Testing
PROCEDURES
NUMBER TITLE REVISION
STS KJ-001A Integrated Diesel Generator and Safeguards Actuation Test 48A
Train A
STS KJ-001B Integrated Diesel Generator and Safeguards Actuation Test 47A
Train B
STS KJ-005A Manual/Auto Start, Synch, & Loading of EDG NE01 58
STN AL-100C TDAFW Pump Reference Pump Curve Determination 1
STS IC-201A Channel Operational Test of Tavg, T and Pressurizer 18
Pressure Protection Set One
STS AB-201B TDAFP Steam Isolation Inservice Valve Test 8
STS AL-201C Turbine Driven Auxiliary Feedwater System Inservice Valve 8
Test
DRAWINGS
NUMBER TITLE REVISION
J-14K81 Compressed Air System Auxiliary Building Details 3
M-12KA05 Piping and Instrumentation Diagram Compressed Air 7
System
M-12AL01 Piping and Instrumentation Diagram Auxiliary Feedwater 23
System
CONDITION REPORTS
00068192 00068267 00068271 00068274 00067526
00060210
WORK ORDER
12-361011-000
A1-15
Section 2RS2: Occupational ALARA Planning and Controls
PROCEDURES
NUMBER TITLE REVISION
AI 16C-008 Work Order Implementation 20A
AI 16C-012 Refuel Preparation & Walk down Guidelines 5
AP 05-009 ALARA Design Guidelines 2A
AP 16C-006 MPAC Work Request / Work Order Process Controls 21
RADIATION WORK PERMITS
NUMBER TITLE REVISION
131000 Health Physics Rover Coverage for RF-19s 2
132001 Mechanical Maintenance Welding Department RWP 5
132002 Maintenance Expanded Scope 6
AUDITS, SELF-ASSESSMENTS, AND SURVEILLANCES
NUMBER TITLE DATE
12-03-RP/PC Radiation Protection/Process Control Programs May 4, 2012
CORRECTIVE ACTION DOCUMENT NUMBERS
64705 65144 65145 65148 65422
65517
WORK ORDERS
09-316582-014 09-346910-000 11-346910-006
A1-16
MISCELLANEOUS DOCUMENTS
NUMBER TITLE DATE
Official Dose for 2010 April 13, 2013
Official Dose for 2011 April 13, 2013
Official Dose for 2012 April 13, 2013
Three Year Rolling Average April 16, 2013
Wolf Creek ALARA Long Range Exposure /Source Term January 10, 2012
Reduction Plan for 2011 - 2016
131000 ALARA Review June, 27, 2012
131000 Post Job ALARA Review April 17, 2013
131000 RWP Budget Report April 17, 2013
132001 ALARA Review December 10,
2012
132001 Post Job ALARA Review May 16, 2013
132001 RWP Budget Report April 18, 2013
132002 ALARA Review February 27, 2013
132002 Post Job ALARA Review April 22, 2013
132002 RWP Budget Report April 15, 2013
Section 2RS4: Occupational Dose Assessment
PROCEDURES
NUMBER TITLE REVISION
RPP 03-121 Neutron Dose Calculations 5
RPP 03-122 Skin Dose Calculations 12
RPP 03-205 DAC-Hour Tracking 16
RPP 03-210 Internal Exposure Calculations and Evaluations 14A
RPP 03-215 Collection of Bioassay Samples 5
RPP 03-406 HP Dosimetry/Records 9
RPP 03-407 Testing of Portal Monitors as Passive Whole Body Counters 1A
A1-17
AUDITS, SELF-ASSESSMENTS, AND SURVEILLANCES
NUMBER TITLE DATE
12-03-RP/PC Radiation Protection / Process Control Program May 4, 2012
CONDITION REPORTS
64625 65544 65925
MISCELLANEOUS DOCUMENTS
NUMBER TITLE REVISION / DATE
APF 30E-OOR- Site Access Training Site Specific - Lesson Plan 6
01
List of RWP Tasks with Multi-Packs May 23, 2013
List of Multipacks per RCA Entry May 23, 2013
Form RPF 03-406 2
Section 4OA1: Performance Indicator Verification
PROCEDURE
NUMBER TITLE REVISION
NEI 99-02 Reactor Oversight Process Performance Indicators 9
Section 4OA2: Identification and Resolution of Problems
CONDITION REPORT
00062234 00025515 00046239 43270 000543
00064461 00064464 00065305 00065418 00065421
00065430 00065799
MISCELLANEOUS
NUMBER TITLE DATE
Wolf Creek Generating Station: Station Roll-up February 4,
Performance Results, 4th Quarter 2012 2013
Wolf Creek Generating Station: Station Roll-up May 6, 2013
Performance Results, 1st Quarter 2013
A1-18
Section 4OA3: Event Follow-Up
PROCEDURES
NUMBER TITLE REVISION
MPM M018Q-01 Standby Diesel Generator Inspection 20
SYS KJ-124 Post Maintenance Run of Emergency Diesel Generator B 52
STS KJ-001B Integrated D/G and Safeguards Actuation Test - Train B 47
STS AC-001 Main Turbine Valve Cycle Test 36
TMP 12-016 Post Modification Main Turbine Control System Generator 7
Startup and Testing
AI 28A-010 Screening Condition Reports 15
AP 05-005 Design, Implementation & Configuration Control of 19
Modifications
AP 19E-002 Reactivity Management Program 17
AP 21-001 Conduct of Operations 61
PROCEDURES
NUMBER TITLE DATE
13-0449 CKL ZL-005A: A EDG Operating Log Rev 4 March 3,
2013
13-0471 CKL ZL-005B: B EDG Operating Log Rev 5 March 3,
2013
13-0567 SYS KJ-123: Post Maintenance Run of Emergency Diesel March 4,
Generator A Rev 53 2013
13-0568 SYS KJ-124: Post Maintenance Run of Emergency Diesel March 4,
Generator B Rev 53 2013
13-0570 MPM M018Q-01: Standby Diesel Generator Inspection Rev March 7,
22 2013
13-0653 SYS KJ-121: Diesel Generator NE01 and NE02 Lineup for March 3,
Automatic Operation Rev 46 2013
CONDITION REPORTS
00064828 65624 67538 67528 67582
67623 70814 68711 68720 68556
70789
A1-19
WORK ORDER
13-365878-002
MISCELLANEOUS
NUMBER TITLE REVISION /
DATE
EN# 48802 NOUE: both diesel generators unavailable 3/1/2013
2009-005 Licensee Event Report: Loss of both Diesel Generators 01
with all Fuel in the Spent Fuel Pool
2013-003 Licensee Event Report: Movement of Irradiated Fuel 00
Progressed After Non-Conservative Decision Making
Resulted in Removal of One Source Range Monitor From
Service
2013-005 Licensee Event Report: Fatigue Failure of Jacket Water 00
Pressure Switch Diaphragm Results in Loss of the B Diesel
Generator
WOL-52624 Failure Analysis of SOR Pressure Switch Manufacturer: April 10,
SOR, Model: 4N6-B5-NX-01A-JJTTX12 Purchase Order 2013
No.: 764426, Exelon Power Labs
Control Room Log March 13,
2013
M-018-01387 Vendor Manual: Installation Operating and Maintenance September
W03 Instructions for Model 20 Air Volume Booster 27, 2005
Fire Event Investigation Report: Auxiliary Boiler Roof April 11, 2013
adjacent to exhaust stack penetration
EN# 48914 Fire Lasting Greater Than 15 Minutes April 11, 2013
DESIGN AND LICENSEE BASIS DOCUMENTS
NUMBER TITLE REVISION
013898 PFSSD - EFHV0060 Control Wiring Modification 0
A1-20
DRAWINGS
NUMBER TITLE REVISION
WIP-E-13EF04A- ESW From Component Cooling Water Heat Exch. Iso. Valve 00
000-A-1, Sh 1 EFHV0060
WIP-E-025- ESW B Return from CCW Heat Exchanger B 0
00007-W13-
A-199
M-12KJ01 Piping & Instrumentation Diagram Standby Diesel 12
Generator Cooling Water System
M-12KJ04 Piping & Instrumentation Diagram Standby Diesel 16
Generator B Cooling Water System
Section 4OA5: Other Activities
WORK ORDERS
12-356794-001 12-356794-003 12-356794-008 12-356794-012
MISCELLANEOUS
NUMBER TITLE DATE
ET 12-0015 Wolf Creek Letter from J. Broschak to U.S. NRC, Re: July 2, 2012
Seismic Aspects of Recommendation 2.3 of the Near-Term
Task Force Review of the Fukushima Dai-ichi Accident
ET 12-0031 Wolf Creek Letter from J. Broschak to U.S. NRC, Re: 180 November
day response to Recommendation 2.3 of the Near-Term 27, 2012
Task Force Review of the Fukushima Dai-ichi Accident
EPRI 1025286 Seismic Walkdown Guidance June 2012
A1-21
Request for Information for Inservice Inspection
Wolf Creek Nuclear Power Plant
February 11, 2013, through February 22, 2013
NRC Inspection Report 05000482/2013002
Please provide the requested information. Thank you for your support.
NOTE: In an effort to keep the requested information organized, please submit the
information using the same request designation. For example, the names and
phone numbers for the program leads should be in a file/folder titled A.5.b.
If you have any questions or comments, please contact the lead inspector Ronald Kopriva at
(817) 200-1104 (Ron.Kopriva@nrc.gov )
PAPERWORK REDUCTION ACT STATEMENT
This letter does not contain new or amended information collection requirements
subject to the Paperwork Reduction Act of 1995 (44 U.S.C. 3501 et seq.).
Existing information collection requirements were approved by the Office of
Management and Budget, control number 3150-0011.
A2-1 Attachment 2
INSERVICE INSPECTION DOCUMENT REQUEST
Inspection Dates: February 11 through February 22, 2013
Inspection Procedures: IP 71111.08 Inservice Inspection (ISI) Activities
Inspectors: Ronald Kopriva, Senior Reactor Inspector (Team Lead)
Megan Williams, Reactor Inspector
A. Information Requested for the In-Office Preparation Week
The following information should be sent to the Region IV office in hard copy or
electronic format (ims.certrec.com preferred), in care of Ronald Kopriva, by February
1, 2013, to facilitate the selection of specific items that will be reviewed during the onsite
inspection week. The inspectors will select specific items from the information requested
below and then request from your staff additional documents needed during the onsite
inspection week (Section B of this enclosure). We ask that the specific items selected
from the lists be available and ready for review on the first day of inspection. Please
provide requested documentation electronically if possible. If requested documents are
large and only hard copy formats are available, please inform the inspector(s), and
provide subject documentation during the first day of the onsite inspection. If you have
any questions regarding this information request, please call the inspector as soon as
possible.
A.1 ISI/Welding Programs and Schedule Information
a) A detailed schedule (including preliminary dates) of:
i) Nondestructive examinations planned for Class 1 & 2 systems and
containment, performed as part of your ASME Section XI, risk informed (if
applicable), and augmented inservice inspection programs during the
upcoming outage.
Provide a status summary of the nondestructive examination inspection
activities vs. the required inspection period percentages for this interval
by category per ASME Section XI, IWX-2400. Do not provide separately
if other documentation requested contains this information.
ii) Reactor pressure vessel head examinations planned for the upcoming
outage.
iii) Examinations planned for Alloy 82/182/600 components that are not
included in the Section XI scope (If applicable).
iv) Examinations planned as part of your boric acid corrosion control
program (Mode 3 walkdowns, bolted connection walkdowns, etc.).
A2-2
v) Welding activities that are scheduled to be completed during the
upcoming outage (ASME Class 1, 2, or 3 structures, systems, or
components).
b) A copy of ASME Section XI Code Relief Requests and associated NRC safety
evaluations applicable to the examinations identified above.
c) A list of nondestructive examination reports (ultrasonic, radiography, magnetic
particle, dye penetrant, Visual VT-1, VT-2, and VT-3), which have identified
relevant conditions on Code Class 1 & 2 systems since the beginning of the last
refueling outage. This should include the previousSection XI pressure test(s)
conducted during start up and any evaluations associated with the results of the
pressure tests. Also, include in the list the nondestructive examination reports
with relevant conditions in the reactor pressure vessel head penetration nozzles
that have been accepted for continued service. The list of nondestructive
examination reports should include a brief description of the structures, systems,
or components where the relevant condition was identified.
d) A list with a brief description (e.g., system, material, pipe size, weld number, and
nondestructive examinations performed) of the welds in Code Class 1 and 2
systems which have been fabricated due to component repair/replacement
activities since the beginning of the last refueling outage, or are planned to be
fabricated this refueling outage.
e) If reactor vessel weld examinations required by the ASME Code are scheduled to
occur during the upcoming outage, provide a detailed description of the welds to
be examined and the extent of the planned examination. Please also provide
reference numbers for applicable procedures that will be used to conduct these
examinations.
f) Copy of any 10 CFR Part 21 reports applicable to your structures, systems, or
components within the scope of Section XI of the ASME Code that have been
identified since the beginning of the last refueling outage.
g) A list of any temporary noncode repairs in service (e.g., pinhole leaks).
h) Please provide copies of the most recent self-assessments for the inservice
inspection, welding, and Alloy 600 programs.
A.2 Reactor Pressure Vessel Head (RPVH)
a) Provide the detailed scope of the planned nondestructive examinations of the
reactor vessel head which identifies the types of nondestructive examination
methods to be used on each specific part of the vessel head to fulfill
commitments made in response to NRC Bulletin 2002-02 and
NRC Order EA-03-009. Also, include examination scope expansion criteria and
planned expansion sample sizes if relevant conditions are identified. (If
applicable)
A2-3
b) A list of the standards and/or requirements that will be used to evaluate
indications identified during nondestructive examination of the reactor vessel
head (e.g., the specific industry or procedural standards which will be used to
evaluate potential leakage and/or flaw indications).
A.3 Boric Acid Corrosion Control Program
a) Copy of the procedures that govern the scope, equipment and implementation of
the inspections required to identify boric acid leakage and the procedures for
boric acid leakage/corrosion evaluation.
b) Please provide a list of leaks (including Code class of the components) that have
been identified since the last refueling outage and associated corrective action
documentation. If during the last cycle, the unit was shutdown, please provide
documentation of containment walkdown inspections performed as part of the
boric acid corrosion control program.
c) Please provide a copy of the most recent self-assessment performed for the
boric acid corrosion control program.
A.4 Steam Generator Tube Inspections
a) A detailed schedule of:
i) Steam generator tube inspection, data analyses, and repair activities for
the upcoming outage (If occurring).
ii) Steam generator secondary side inspection activities for the upcoming
outage. (If occurring).
b) Please provide a copy of your steam generator inservice inspection program and
plan. Please include a copy of the operational assessment from last outage and
a copy of the following documents as they become available:
i) Degradation assessment
ii) Condition monitoring assessment
c) If you are planning on modifying your Technical Specifications such that they are
consistent with Technical Specification Task Force Traveler TSTF-449, Steam
Generator Tube Integrity, please provide copies of your correspondence with the
NRC regarding deviations from the standard technical specifications.
d) Copy of steam generator history documentation given to vendors performing
eddy current testing of the steam generators during the upcoming outage.
A2-4
e) Copy of steam generator eddy current data analyst guidelines and site validated
eddy current technique specification sheets. Additionally, please provide a copy
of EPRI Appendix H, Examination Technique Specification Sheets, qualification
records.
f) Identify and quantify any steam generator tube leakage experienced during the
previous operating cycle. Also provide documentation identifying which steam
generator was leaking and corrective actions completed or planned for this
condition (If applicable).
g) Provide past history of the condition and issues pertaining to the secondary side
of the steam generators (including items such as loose parts, fouling, top of tube
sheet condition, crud removal amounts, etc.)
h) Provide copies of your most recent self assessments of the steam generator
monitoring, loose parts monitoring, and secondary side water chemistry control
programs.
i) Indicate where the primary, secondary, and resolution analyses are scheduled to
take place.
j) Provide a summary of the scope of the steam generator tube examinations,
including examination methods such as Bobbin, Rotating Pancake, or Plus Point,
and the percentage of tubes to be examined. Do not provide these documents
separately if already included in other information requested.
A.5 Additional Information Related to all Inservice Inspection Activities
a) A list with a brief description of inservice inspection, boric acid corrosion control
program, and steam generator tube inspection related issues (e.g., condition
reports) entered into your corrective action program since the beginning of the
last refueling outage (for Unit 2). For example, a list based upon data base
searches using key words related to piping or steam generator tube degradation
such as: inservice inspection, ASME Code,Section XI, NDE, cracks, wear,
thinning, leakage, rust, corrosion, boric acid, or errors in piping/steam generator
tube examinations.
b) Please provide names and phone numbers for the following program leads:
Inservice inspection (examination, planning)
Containment exams
Reactor pressure vessel head exams
Snubbers and supports
Repair and replacement program
Licensing
Site welding engineer
Boric acid corrosion control program
A2-5
Steam generator inspection activities (site lead and vendor contact)
B. Information to be Provided Onsite to the Inspector(s) at the Entrance Meeting
(February 11, 2013):
B.1 Inservice Inspection / Welding Programs and Schedule Information
a) Updated schedules for inservice inspection/nondestructive examination activities,
including steam generator tube inspections, planned welding activities, and
schedule showing contingency repair plans, if available.
b) For ASME Code Class 1 and 2 welds selected by the inspector from the lists
provided from section A of this enclosure, please provide copies of the following
documentation for each subject weld:
i) Weld data sheet (traveler)
ii) Weld configuration and system location
iii) Applicable Code Edition and Addenda for weldment
iv) Applicable Code Edition and Addenda for welding procedures
v) Applicable weld procedures used to fabricate the welds
vi) Copies of procedure qualification records supporting the weld procedures
from B.1.b.v
vii) Copies of mechanical test reports identified in the procedure qualification
records above
viii) Copies of the nonconformance reports for the selected welds (If
applicable)
ix) Radiographs of the selected welds and access to equipment to allow
viewing radiographs (If radiographic testing was performed)
x) Copies of the preservice examination records for the selected welds
xi) Copies of welder performance qualifications records applicable to the
selected welds, including documentation that welder maintained
proficiency in the applicable welding processes specified in the weld
procedures (at least 6 months prior to the date of subject work)
xii) Copies of nondestructive examination personnel qualifications (Visual
inspection, penetrant testing, ultrasonic testing, radiographic testing), as
applicable
c) For the inservice inspection related corrective action issues selected by the
inspectors from section A of this enclosure, provide a copy of the corrective
actions and supporting documentation.
d) For the nondestructive examination reports with relevant conditions on Code
Class 1 and 2 systems selected by the inspectors from Section A above, provide
a copy of the examination records, examiner qualification records, and
associated corrective action documents.
A2-6
e) A copy of (or ready access to) most current revision of the inservice inspection
program manual and plan for the current Interval.
f) For the nondestructive examinations selected by the inspectors from section A of
this enclosure, provide a copy of the nondestructive examination procedures
used to perform the examinations (including calibration and flaw
characterization/sizing procedures). For ultrasonic examination procedures
qualified in accordance with ASME Section XI, Appendix VIII, provide
documentation supporting the procedure qualification (e.g., the EPRI
performance demonstration qualification summary sheets). Also, include
qualification documentation of the specific equipment to be used (e.g., ultrasonic
unit, cables, and transducers including serial numbers) and nondestructive
examination personnel qualification records.
B.2 Reactor Pressure Vessel Head
a) Provide the nondestructive personnel qualification records for the examiners who
will perform examinations of the reactor pressure vessel head.
b) Provide drawings showing the following: (If a visual examination is planned for
the upcoming refueling outage)
i) Reactor pressure vessel head and control rod drive mechanism nozzle
configurations
ii) Reactor pressure vessel head insulation configuration
Note: The drawings listed above should include fabrication drawings for
the nozzle attachment welds as applicable.
c) Copy of nondestructive examination reports from the last reactor pressure vessel
head examination.
d) Copy of evaluation or calculation demonstrating that the scope of the visual
examination of the upper head will meet the 95 percent minimum coverage
required by NRC Order EA-03-009 (If a visual examination is planned for the
upcoming refueling outage).
e) Provide a copy of the procedures that will be used to identify the source of any
boric acid deposits identified on the reactor pressure vessel head. If no explicit
procedures exist which govern this activity, provide a description of the process
to be followed including personnel responsibilities and expectations.
f) Provide a copy of the updated calculation of effective degradation years for the
reactor pressure vessel head susceptibility ranking.
g) Provide copy of the vendor qualification report(s) that demonstrates the detection
capability of the nondestructive examination equipment used for the reactor
pressure vessel head examinations. Also, identify any changes in equipment
A2-7
configurations used for the reactor pressure vessel head examinations which
differ from that used in the vendor qualification report(s).
B.3 Boric Acid Corrosion Control Program
a) Please provide boric acid walkdown inspection results, an updated list of boric
acid leaks identified so far this outage, associated corrective action
documentation, and overall status of planned boric acid inspections.
b) Please provide any engineering evaluations completed for boric acid leaks
identified since the end of the last refueling outage. Please include a status of
corrective actions to repair and/or clean these boric acid leaks. Please identify
specifically which known leaks, if any, have remained in service or will remain in
service as active leaks.
B.4 Steam Generator Tube Inspections
a) Copies of the Examination Technique Specification Sheets and associated
justification for any revisions.
b) Copy of the guidance to be followed if a loose part or foreign material is identified
in the steam generators.
c) Please provide a copy of the eddy current testing procedures used to perform the
steam generator tube inspections (specifically calibration and flaw
characterization/sizing procedures, etc.). Also include documentation for the
specific equipment to be used.
d) Please provide copies of your responses to NRC and industry operating
experience communications such as Generic Letters, Information Notices, etc.
(as applicable to steam generator tube inspections) Do not provide these
documents separately if already included in other information requested such as
the degradation assessment.
e) List of corrective action documents generated by the vendor and/or site with
respect to steam generator inspection activities.
B.5 Codes and Standards
a) Ready access to (i.e., copies provided to the inspector(s) for use during the
inspection at the onsite inspection location, or room number and location where
available):
i) Applicable Editions of the ASME Code (Sections V, IX, and XI) for the
inservice inspection program and the repair/replacement program.
ii) EPRI and industry standards referenced in the procedures used to
perform the steam generator tube eddy current examination.
A2-8
Inspector Contact Information:
Ronald Kopriva
Senior Reactor Inspector
817-200-1104
Ron.Kopriva@nrc.gov
Mailing Address:
US NRC Region IV
Attn: Ronald Kopriva
1600 E. Lamar Blvd
Arlington, TX 76011
A2-9
The following items are requested for the
Occupational Radiation Safety Inspection
at Wolf Creek Generating Plant
May 20-24, 2013
Integrated Report 2013003
Inspection areas are listed in the attachments below.
Please provide the requested information on or before May 3, 2013.
Please submit this information using the same lettering system as below. For example, all
contacts and phone numbers for Inspection Procedure 71124.01 should be in a file/folder titled
1- A, applicable organization charts in file/folder 1- B, etc.
If information is placed on ims.certrec.com, please ensure the inspection exit date entered is at
least 30 days later than the onsite inspection dates, so the inspectors will have access to the
information while writing the report.
In addition to the corrective action document lists provided for each inspection procedure listed
below, please provide updated lists of corrective action documents at the entrance meeting.
The dates for these lists should range from the end dates of the original lists to the day of the
entrance meeting.
If more than one inspection procedure is to be conducted and the information requests appear
to be redundant, there is no need to provide duplicate copies. Enter a note explaining in which
file the information can be found.
If you have any questions or comments, please contact Larry Ricketson at (817) 200-1165 or
Larry.Ricketson@nrc.gov.
PAPERWORK REDUCTION ACT STATEMENT
This letter does not contain new or amended information collection requirements subject
to the Paperwork Reduction Act of 1995 (44 U.S.C. 3501 et seq.). Existing information
collection requirements were approved by the Office of Management and Budget,
control number 3150-0011.
A3-1 Attachment 3
2. Occupational ALARA Planning and Controls (71124.02)
Date of Last Inspection: September 24, 2012
A. List of contacts and telephone numbers for ALARA program personnel
B. Applicable organization charts
C. Copies of audits, self-assessments, and LERs, written since date of last inspection,
focusing on ALARA
D. Procedure index for ALARA Program
E. Please provide specific procedures related to the following areas noted below.
Additional Specific Procedures may be requested by number after the inspector reviews
the procedure indexes.
1. ALARA Program
2. ALARA Committee
3. Radiation Work Permit Preparation
F. A summary list of corrective action documents (including corporate and subtiered
systems) written since date of last inspection, related to the ALARA program. In addition
to ALARA, the summary should also address Radiation Work Permit violations,
Electronic Dosimeter Alarms, and RWP Dose Estimates
NOTE: The lists should indicate the significance level of each issue and the search
criteria used. Please provide documents which are searchable.
G. List of work activities greater than 1 rem, since date of last inspection.
Include original dose estimate and actual dose.
H. Site dose totals and 3-year rolling averages for the past 3 years (based on dose of
record)
I. Outline of source term reduction strategy
A3-2
4. Occupational Dose Assessment (Inspection Procedure 71124.04)
Date of Last Inspection: August 15, 2011
A. List of contacts and telephone numbers for the following areas:
1. Dose Assessment personnel
B. Applicable organization charts
C. Audits, self assessments, vendor or NUPIC audits of contractor support, and LERs
written since date of last inspection, related to:
1. Occupational Dose Assessment
D. Procedure indexes for the following areas
1. Occupational Dose Assessment
E. Please provide specific procedures related to the following areas noted below.
Additional Specific Procedures will be requested by number after the inspector reviews
the procedure indexes.
1. Radiation Protection Program
2. Radiation Protection Conduct of Operations
3. Personnel Dosimetry Program
4. Radiological Posting and Warning Devices
5. Air Sample Analysis
6. Performance of High Exposure Work
7. Declared Pregnant Worker
8. Bioassay Program
F. List of corrective action documents (including corporate and subtiered systems) written
since date of last inspection, associated with:
1. NVLAP accreditation
2. Dosimetry (TLD/OSL, etc.) problems
3. Electronic alarming dosimeters
4. Bioassays or internally deposited radionuclides or internal dose
5. Neutron dose
NOTE: The lists should indicate the significance level of each issue and the search
criteria used.
G. List of positive whole body counts since date of last inspection, names redacted if
desired
H. Part 61 analyses/scaling factors
A3-3