ML112240223
| ML112240223 | |
| Person / Time | |
|---|---|
| Site: | Wolf Creek  |
| Issue date: | 08/12/2011 |
| From: | Geoffrey Miller NRC/RGN-IV/DRP/RPB-B |
| To: | Matthew Sunseri Wolf Creek |
| References | |
| EA-11-149 IR-11-003 | |
| Download: ML112240223 (103) | |
See also: IR 05000482/2011003
Text
UNITED STATES
NUCLEAR REGULATORY COMMISSION
REGI ON I V
612 EAST LAMAR BLVD, SUITE 400
ARLINGTON, TEXAS 76011-4125
August 12, 2011
Matthew Sunseri, President and
Chief Executive Officer
Wolf Creek Nuclear Operating Corporation
P.O. Box 411
Burlington, KS 66839
Subject: WOLF CREEK GENERATING STATION - NRC INTEGRATED INSPECTION
REPORT AND NOTICE OF VIOLATION 05000482/2011003
Dear Mr. Sunseri:
On June 30, 2011, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at
your Wolf Creek Generating Station. The enclosed integrated inspection report documents the
inspection findings, which were discussed on July 13, 2011, with Mr. Stephen Hedges, Site Vice
President, and other members of your staff.
The inspections examined activities conducted under your license as they relate to safety and
compliance with the Commissions rules and regulations and with the conditions of your license.
The inspectors reviewed selected procedures and records, observed activities, and interviewed
personnel.
Based on the results of this inspection, one violation is cited in the enclosed Notice of Violation
(Notice) and the circumstances surrounding this violation are described in detail in the enclosed
report. The violation involved the failure to implement procedures for opening of main steam
isolation valves without causing safety system actuations (EA-11-149). Although determined to
be of very low safety significance (Green), this violation is being cited in the Notice because
Wolf Creek failed to restore compliance within a reasonable time after the violation was
identified in NRC Inspection Report 05000482/2010004, per Section 2.3.2 of 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.
Please note you are required to respond to this letter and should follow the instructions specified
in the enclosed Notice. If you have additional information that you believe the NRC should
consider, you may provide it in your response to the Notice. The NRC will use your response, in
part, to determine whether further enforcement action is necessary to ensure compliance with
regulatory requirements.
This report also documents nine additional NRC-identified and self-revealing issues that were
evaluated under the risk significance determination process as having very low safety
Wolf Creek Nuclear Operating Corporation -2-
significance (Green). The NRC determined that violations are associated with eight of these
issues. Additionally, two licensee-identified violations, which were determined to be of very low
safety significance, are listed in this report. However, because of the very low safety
significance and because they were entered into your corrective action program, the NRC is
treating these findings as noncited violations, consistent with Section 2.3.2 of the NRC
If you contest the violation or the significance of the noncited violations, you should provide a
response within 30 days of the date of this inspection report, with the basis for your denial, to
the U.S. Nuclear Regulatory Commission, ATTN: Document Control Desk, Washington, D.C.
20555-0001, with copies to the Regional Administrator, U.S. Nuclear Regulatory Commission,
Region IV, 612 E. Lamar Blvd, Suite 400, Arlington, Texas, 76011-4125; the Director, Office of
Enforcement, U.S. Nuclear Regulatory Commission, Washington, D.C. 20555-0001; and the
NRC Resident Inspector at the facility. In addition, if you disagree with the cross-cutting aspect
assigned to any finding 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 the facility.
In accordance with 10 CFR 2.390 of the NRC's "Rules of Practice," a copy of this letter, its
enclosure, and your response, if you choose to provide one for cases where a response is not
required, will be made available electronically for public inspection in the NRC Public Document
Room or from the NRC's document system (ADAMS), accessible from the NRC Web site at
http://www.nrc.gov/reading-rm/adams.html. 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/
Geoffrey B. Miller, Chief
Project Branch B
Division of Reactor Projects
Docket No. 50-482
License No. NPF-42
Enclosure:
NRC Inspection Report and Notice of Violation 05000482/2011003
w/Attachment: Supplemental Information
cc w/Enclosure:
Distribution via Listserv
Electronic distribution by RIV:
Wolf Creek Nuclear Operating Corporation -3-
Regional Administrator (Elmo.Collins@nrc.gov)
Deputy Regional Administrator (Art.Howell@nrc.gov)
DRP Director (Kriss.Kennedy@nrc.gov)
DRP Deputy Director (Jeff.Clark@nrc.gov)
DRS Director (Anton.Vegel@nrc.gov)
Acting DRS Director (Robert.Calwell@nrc.gov)
DRS Deputy Director (Tom.Blount@nrc.gov)
Senior Resident Inspector (Chris.Long@nrc.gov)
Resident Inspector (Charles.Peabody@nrc.gov)
WC Administrative Assistant (Shirley.Allen@nrc.gov)
Branch Chief, DRP/B (Geoffrey.Miller@nrc.gov)
Senior Project Engineer, DRP/B (Rick.Deese@nrc.gov)
Senior Project Engineer, DRP/B (Leonard.Willoughby@nrc.gov)
Project Engineer, DRP/B (Nestor.Makris@nrc.gov)
Public Affairs Officer (Victor.Dricks@nrc.gov)
Public Affairs Officer (Lara.Uselding@nrc.gov)
Project Manager (Randy.Hall@nrc.gov)
Branch Chief, DRS/TSB (Michael.Hay@nrc.gov)
RITS Coordinator (Marisa.Herrera@nrc.gov)
Regional Counsel (Karla.Fuller@nrc.gov)
Congressional Affairs Officer (Jenny.Weil@nrc.gov)
OEMail Resource
ROPreports
RIV/ETA/OEDO (John.McHale@nrc.gov)
DRS/TSB STA (Dale.Powers@nrc.gov)
R:\_REACTORS\_WC\2011\WC2011003-RP-CML
ADAMS: No Yes SUNSI Review Complete Reviewer Initials: RWD
Publicly Available Non-Sensitive
Non-publicly Available Sensitive
SRI:DRP/B RI:DRP/B C:DRS/EB1 C:DRS/EB2 DRS/PSB1
CLong CPeabody TFarnholtz NOKeefe MHay
/E-GBM/ /E-GBM/ /RA/ /JMateychick for/ /JLarson for/
8/12/2011 8/2/2011 8/9/2011 8/9/2011 8/10/2011
C:DRS/OB C:DRS/TSB DRS/PSB2 RIV:ACES GMiller
MHaire DPowers GWerner RKellar
/RA/ /RA/ /RA/ /RA/ /RA/
8/10/2011 8/10/2011 8/9/2011 8/11/2011 8/12/2011
OFFICIAL RECORD COPY T=Telephone E=E-mail F=Fax
NOTICE OF VIOLATION
Wolf Creek Nuclear Operating Corporation Docket: 50-482
Wolf Creek Generating Station License No: NPF-42
During an NRC inspection conducted March 19 through June 30, 2011 a violation of an NRC
requirement was identified. In accordance with the NRC Enforcement Policy, the violation is
listed below:
Technical Specification 5.4.1.a requires that procedures be established,
implemented, and maintained covering the activities described in Regulatory
Guide 1.33, Revision 2, Appendix A. Regulatory Guide 1.33, Revision 2,
Appendix A, Section 3.i requires procedures for the startup, operation and shutdown
of the main steam system. Wolf Creek Procedure SYS AB-120, Main Steam and
Steam Dump Startup and Operation, Revision 27, implements these requirements
for the main steam system.
Contrary to the above, from March 5, 2010, to March 19, 2011, Wolf Creek
Procedure SYS AB-120 had not been maintained to cover activities for the startup,
operation and shutdown of the main steam system. Specifically,
Procedure SYS AB-120, Revision 27, contained inadequate steps necessary to open
a main steam isolation valve without causing a safety injection signal.
This violation is associated with a Green Significance Determination Process finding
(EA-11-149).
Pursuant to the provisions of 10 CFR 2.201, Wolf Creek Nuclear Operating Corporation is
required to submit a written statement or explanation to the U.S. Nuclear Regulatory
Commission, ATTN: Document Control Desk, Washington, DC 20555, with a copy to the
Regional Administrator, Region IV, and a copy to the NRC Senior Resident Inspector at the
facility that is the subject of this Notice of Violation (Notice), within 30 days of the date of the
letter transmitting this Notice. This reply should be clearly marked as a "Reply to Notice of
Violation EA-11-149," 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 to avoid
further violations, 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.
-1- Enclosure
Because your response will be made available electronically for public inspection in the NRC
Public Document Room or from the NRCs document 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.
Dated this 12th day of August 2011.
-2- Enclosure
U.S. NUCLEAR REGULATORY COMMISSION
REGION IV
Docket: 05000482
License: NPF-42
Report: 05000482/2011003
Licensee: Wolf Creek Nuclear Operating Corporation
Facility: Wolf Creek Generating Station
Location: 1550 Oxen Lane NE
Burlington, Kansas
Dates: April 1 to June 30, 2011
Inspectors: C. Long, Senior Resident Inspector
C. Peabody, Resident Inspector
D. Reinert, Acting Resident Inspector
J. Drake, Senior Reactor Inspector
A. Fairbanks, Reactor Inspector
G. Guerra, CHP, Emergency Preparedness Inspector
G. Pick, Senior Reactor Inspector
D. Strickland, Operations Engineer
Approved By: G. Miller, Chief, Project Branch B
Division of Reactor Projects
-3- Enclosure
SUMMARY OF FINDINGS
IR 05000482/2011003, 4/1 - 6/30/2011; Wolf Creek Generating Station, Integrated Resident
Report, Adverse Weather Protection, Equipment Alignments, Inservice Inspection Activities,
Postmaintenance Testing, Event Follow-up, and Other Activities.
The report covered a 3-month period of inspection by resident inspectors and announced
baseline inspections by region-based inspectors. One Green cited violation, eight Green
noncited violations, and one finding of significance were 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 noncited violation of Technical
Specification 5.4.1.a, Administrative Procedures, for having no procedure to
address onsite debris impacting plant equipment during severe weather. The
inspectors walked down external areas of the plant on June 1 and June 9, 2011,
prior to the onset of predicted severe thunderstorms and tornadoes. The
inspectors found loose debris each time and brought it to the attention of the
licensee who secured the materials. The inspectors walked down the
transformer yard and tank yard during a thunderstorm on June 16 and found
loose debris such as plywood, trash, wood planks, and fiberglass planks. The
inspectors brought this to the attention of Wolf Creek and the materials were
removed or secured. Wolf Creek initiated several condition reports but they only
addressed immediate cleanup. Wolf Creek procedures had no steps for securing
potential wind-driven projectiles prior to severe weather. After June 16, Wolf
Creek wrote Condition Report 40573 which started a weekly maintenance activity
to remove loose materials and added procedure steps to have operations walk
down external areas prior to severe weather.
This finding was more than minor because it impacted the protection against
external factors attribute of the Initiating Events Cornerstone, and it 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. The inspectors evaluated this finding using Inspection Manual
Chapter 0609.04, and determined that it was of very low safety significance
(Green) for June 16, 2011, because it did not contribute to both the likelihood of a
reactor trip and the likelihood that mitigation equipment would be unavailable
since the reactor was shutdown. Inspectors used Manual Chapter 0609
Appendix G, Checklist 4 for the other occurrences because Wolf Creek was in
-4- Enclosure
Modes 4 or 5. The finding again screened to Green because it did not increase
the likelihood of a loss of inventory, did not cause the loss of reactor coolant
system instrumentation, did not degrade the ability of the licensee to terminate a
leak path or add inventory when needed, or degrade the ability to recover
residual heat removal if it was lost. This finding has a cross-cutting aspect in the
area of problem identification and resolution, specifically the corrective action
program attribute because licensees short-term corrective actions failed to
ensure debris was secured or removed prior to severe weather
P.1(d)(Section 1R01).
- Green. The inspectors documented a self-revealing noncited violation of 10 CFR
Part 50, Appendix B, Criterion IX, Control of Special Processes. Specifically, in
October 2009, welders failed to ensure the fillet weld between the train B
charging header and the half coupling used to attach two vent valves met the
specified weld requirements. This weld failed in January 2011, rendering the
train B charging system inoperable. The licensees extent of condition review
identified 12 vent line welds which did not meet ASME code weld size
requirements and/or procedural requirements for 2:1 weld taper configuration.
Additionally, quality assurance inspectors failed to identify that the 2:1 taper weld
requirements specified by procedure, and ASME minimum weld size
requirements, were not met in multiple vent line welds. The weld was repaired
and built up to the correct 2:1 aspect ratio. This issue was entered into the
licensees corrective action program as Condition Reports 32648, 33686, 33689,
and 36438.
The finding was more than minor because it was associated with the equipment
performance attribute of the Initiating Events Cornerstone and adversely affected
the cornerstone objective to limit the likelihood of those events that upset plant
stability and challenge critical safety functions during power operations. The
inspectors performed a Phase 1 screening in accordance with Inspection Manual
Chapter 0609.04, Phase 1 - Initial Screening and Characterization of Findings,
and determined that the finding was of very low safety significance (Green)
because the issue did not result in exceeding the technical specification limit for
identified reactor coolant system leakage or affect other mitigating systems
resulting in a total loss of their safety function. This finding had a cross-cutting
aspect in the area of human performance, resources, because the licensee failed
to ensure that personnel, specifically welders and quality assurance inspectors,
were adequately trained in the procedural requirements and methods for
measuring weld dimensions to assure nuclear safety H.2(b)(Section 1R08).
- Green. The inspectors identified a noncited violation of 10 CFR Part 50 involving
the failure of the licensee to ensure that weld preparation was protected from
deleterious contamination in that drawers (located in the hot tool room)
containing files, grinding wheels, flapper wheels, and cutting wheels, used for the
purpose of weld preparation, contained a mixture of both stainless steel tools and
carbon steel tools. The failure to separate tools used for stainless steel weld
preparation from tools used for carbon steel preparation could result in the
-5- Enclosure
contamination of stainless steel welds by carbon steel and affect the material
integrity and corrosion resistance. The licensee immediately removed the tools
and replaced them with new tools stored separately for use on specific types of
metal. This issue was entered into the licensees corrective action program as
Condition Report 36444.
The finding was more than minor because it was associated with the equipment
performance attribute of the Initiating Events Cornerstone and adversely affected
the cornerstone objective to limit the likelihood of those events that upset plant
stability and challenge critical safety functions during power operations, and if left
uncorrected the finding would become a more significant safety concern. The
inspectors performed a Phase 1 screening in accordance with Inspection Manual
Chapter 0609.04, Phase 1 - Initial Screening and Characterization of Findings,
and determined that the finding was of very low safety significance (Green)
because the issue did not result in exceeding the technical specification limit for
identified reactor coolant system leakage or affect other mitigating systems
resulting in a total loss of their safety function. This finding had a cross-cutting
aspect in the area of human performance, resources, because the licensee did
not provide complete, accurate, and up-to-date procedures for the preparation of
stainless steel and carbon steel welds H.2(c)(Section 1R08).
- Green. The inspectors identified a noncited violation of 10 CFR Part 50,
Appendix B, Criterion III, Design Control, involving the failure of the licensee to
review the suitability of installing brass fittings and leaving test fittings on
pressure, differential pressure, and flow transmitter equalizing block valve drain
ports instead of the design specified stainless steel manifold plugs. During a
boric acid walkdown, the inspectors identified that drain ports on the equalizing
block of two separate reactor coolant system flow transmitters had brass fittings
installed instead of the design specified stainless steel fittings. In response to
inspector concerns about the brass fittings, the licensee subsequently discovered
that a design configuration nonconformance existed by leaving the test fittings on
the drain port during plant operation. Licensee Drawing J-17D22 specifies that
manifold plugs be installed in the drain ports during plant operation. The licensee
immediately replaced the brass caps with stainless steel fittings. This issue was
entered into the licensees corrective action program as Condition Report 36439.
The finding was more than minor because it was associated with the design
control attribute of the Initiating Events Cornerstone and adversely affected the
cornerstone objective to limit the likelihood of those events that upset plant
stability and challenge critical safety functions during power operations. The
inspectors performed a Phase 1 screening in accordance with Inspection
Manual 0609.04, Phase 1 - Initial Screening and Characterization of Findings,
and determined that the finding was of very low safety significance (Green)
because the issue would not result in exceeding the technical specification limit
for identified reactor coolant system leakage or affect other mitigating systems
resulting in a total loss of their safety function. The inspectors also determined
that the finding had a cross-cutting aspect in the area of human performance,
resources, because the licensee did not provide adequate training of personnel
-6- Enclosure
so that the inappropriately installed fittings could be identified during system
walkdowns H.2(b)(Section 1R08).
- Green. The inspectors identified a cited violation of Technical
Specification 5.4.1.a, Administrative Procedures, involving Wolf Creeks failure
to correct Procedure SYS AB-120 for main steam isolation valve operation.
Specifically, between March 3, 2010, and March 19, 2011, Wolf Creek
experienced repeat cases of safety-system actuations due to
Procedure SYS AB-120 containing inadequate steps to establish conditions
necessary to open a main steam isolation valve. Corrective actions were
previously limited to steam header pressures below 300 psi. Wolf Creek
commenced a root cause evaluation of the March 19, 2011, safety injection
under Condition Report 34964. Due to Wolf Creeks failure to restore compliance
from previous NCV 05000482/2010004-01 within a reasonable time after the
violation was identified, this violation is being cited as a Notice of Violation
consistent with the Enforcement Policy.
Failure to correct deficiencies in Procedure SYS AB-120 for steam pressures
above 300 psi was a performance deficiency. The inspectors determined that
this finding was more than minor because it impacted the equipment
performance attribute for the Initiating Events Cornerstone, and it affected the
cornerstone objective to limit the likelihood of those events that upset plant
stability and challenge critical safety functions during shutdown as well as power
operations. Specifically, this issue relates to the configuration control attribute for
shut down equipment alignment. The inspectors evaluated the significance of
this finding using Inspection Manual Chapter 0609.04. Assuming worst case
degradation, the finding resulted in exceeding the technical specification limit for
reactor coolant system leakage due to the pressurizer power-operated relief
valve cycling. Therefore, the inspectors screened the finding to a Phase 2 review
by the senior reactor analyst. The senior reactor analyst used the Wolf Creek
SPAR model and concluded that the incremental core damage probability
was 3.7E-7 (Green). The inspectors found that the cause of the finding has a
cross-cutting aspect in the area of problem identification and resolution
associated with the corrective action program. Specifically, several evaluations
failed to have an adequate extent of condition review and did not find that
procedures were inadequate for opening a main steam isolation valve above
300 psi P.1(c)(Section 4OA3.1).
- Green. The inspectors reviewed a self-revealing noncited violation of Technical
Specification 5.4.1.a, Administrative Procedures, for failure to follow procedural
requirements to maintain reactor coolant system pressure below 350 psig.
Control room operators increased charging flow at too great a rate with the
reactor coolant system water-solid which caused the pressurizer power-operated
relief valve to cycle three times over several minutes until adjustments to letdown
could be made to reduce reactor coolant system pressure. Also, the letdown
pressure controller was left in manual when automatic control would have
lessened the pressure increase. Wolf Creek wrote Condition Report 35244 to
-7- Enclosure
correct the deficiency by changing several procedures for water-solid plant
operations.
The failure to maintain pressure below the power-operated relief valve setpoint
was a performance deficiency. The performance deficiency was more than
minor because it impacted the Initiating Events Cornerstone objective of
configuration control to limit the likelihood of those events that upset plant
stability and challenge critical safety functions during shutdown as well as power
operations. The significance of the finding was determined using Inspection
Manual Chapter 0609, Significance Determination Process, Appendix G,
Checklist 2, and determined to be of very low safety significance (Green),
because it did not cause the loss of mitigating capability of core heat removal,
inventory control, power availability, containment control, or reactivity control.
Additionally, the finding also did not cause any low temperature overpressure
technical specifications to be exceeded. The inspectors found that the cause of
the finding had a cross-cutting aspect in the area of human performance.
Specifically, operators had to rely on skill of the craft when procedures should
have supplied more instruction for manipulating charging and letdown with a
water-solid plant H.2.c](Section 4OA3.2).
- Green. The inspectors reviewed a self-revealing noncited violation of License
Condition 2.C.5 for failure to implement adequate fire watches which affected
both trains of vital ac and dc switchgear. The inadequate fire watches occurred
during an actual fire which negated the Halon system discharge because internal
fire doors were not shut, as required, by the fire watch. The inspectors found
problems with fire impairments and watches from 2008 that had not been
corrected. Subsequent to the fire, Wolf Creek again briefed and trained its
personnel on the requirements for fire watches. This issue is captured in the
corrective action program as Condition Report 36719.
Failure to implement adequate fire impairments such that the fire watches
ensured the success of the Halon system was a performance deficiency. The
performance deficiency was more than minor because it impacted the Initiating
Events Cornerstone and its objective to limit the likelihood of those events that
upset plant stability and challenge critical safety functions during shutdown as
well as power operations. Specifically, the protection against external factors
attribute was impacted by the fire impairment. To determine significance, the
inspectors used Inspection Manual Chapter 0609.04 to screen the finding to
Inspection Manual Chapter 0609, Appendix F, because the fire protection
defense-in-depth strategies involving automatic suppression, fire barriers, and
administrative controls were degraded. The senior reactor analyst conducted a
Phase 3 review of this finding and concluded that the incremental core damage
frequency was 1.6E-8 per year, or very low safety significance (Green). The
inspectors found that the cause of the finding had a cross-cutting aspect in the
area of problem identification and resolution. Specifically, corrective actions from
ineffective fire watches in 2008 did not prevent recurrence of the inadequate fire
watch on April 5, 2011 P.1.d](Section 4OA3.3).
-8- Enclosure
Cornerstone: Mitigating Systems
- Green. The inspectors reviewed a self-revealing noncited violation of Technical
Specification 5.4.1a, Administrative Procedures, for a loss of component
cooling water train B inventory caused by inadequate clearance order
verification. Valve HBV110 was stuck in position and was partially open. When
the clearance order was implemented, the operators concluded the valve was
already closed. Subsequently, the valve created a leakage path which exceeded
the surge tank makeup flow capacity and required manual isolation by the control
room operators to protect safety-related components. Wolf Creek has taken
corrective actions to include communication of expected as-found equipment
positions in pre-job briefings and the clearance order template. This issue is
captured in the corrective action program as Condition Reports 34505
and 40219.
Failure to properly establish clearance order boundary isolation was a
performance deficiency. The performance deficiency is more than minor
because it is associated with the equipment performance and human
performance attributes of the Mitigating Systems Cornerstone and impacted the
cornerstone objective of ensuring the availability, reliability, and capability of
systems that respond to initiating events to prevent undesirable consequences.
Using Inspection Manual Chapter 0609.04, the finding was determined to be of
very low safety significance because the finding did not result in the loss of
operability or functionality of the component cooling water train or screen as
potentially risk significant due to a seismic, flooding, or severe weather initiating
event or screen as potentially risk significant due to a seismic, flooding, or severe
weather initiating event. The inspectors found that the finding had a cross-cutting
aspect of work practices in the area of human performance associated with the
communication of human error prevention techniques, such as holding pre-job
briefings, self- and peer-checking, and proper documentation of activities
H.4(a)(Section 1R04).
- Green. The inspectors identified a finding involving the failure to follow the
requirements of Procedure AP 16E-002, Post Maintenance Testing
Development, for the startup feedwater pump. On November 4-6, 2010, Wolf
Creek workers disassembled the startup feedwater pump for numerous
preventive and corrective activities including removing the rotating element. On
November 17, 2010, Wolf Creek conducted surveillance Procedure STN AE-007,
Startup Main Feedwater Pump Operational Test, following reassembly. The
only acceptance criteria listed in this procedure is that the motor-driven feedwater
pump starts from the control room with no local operator action. The inspectors
found this contrary to Procedure AP 16E-002, which requires acceptance criteria
for a pump flow capacity test, vibration, bearing and lubrication temperatures,
motor current, external leakage, and lubrication level be found satisfactory. This
issue is captured in the corrective action program as Condition Report 39494.
Wolf Creek issued a new work package to conduct a single-point pump capacity
test and complete the required postmaintenance testing. Wolf Creek found,
pending final review, that initial calculations show that the pump design is
-9- Enclosure
capable of enough flow to provide a heat sink in emergency operating
procedures.
Failure to follow Procedure AP 16E-002 for developing test criteria for plant
equipment after the completion of maintenance activities is a performance
deficiency. The finding is more than minor because it is associated with the
Mitigating Systems Cornerstone attribute of equipment performance and it
adversely affects the cornerstone objective of ensuring the availability, reliability,
and capability of systems that respond to initiating events to prevent undesirable
consequences. Using Inspection Manual Chapter 0609.04, the inspectors
determined that the finding had very low safety significance (Green) because it
did not result in a loss of system safety function, an actual loss of safety function
of a single train for greater than its technical specification allowed outage time, or
screen as potentially risk significant due to a seismic, flooding, or severe weather
initiating event. The inspectors determined that the finding had a cross-cutting
aspect in the area of problem identification and resolution. Specifically, Wolf
Creek created a testing procedure in response to a root cause evaluation, but did
not consider acceptance criteria to ensure that the pump performs acceptably
P.1(d)(Section 1R19).
- Green. The inspectors identified a noncited violation of 10 CFR Part 50,
Appendix B, Criterion III, Design Control, for the failure to translate the design
basis into instructions, procedures, and drawings. The inspectors found that the
licensee failed to assess whether vortexing occurred in the containment spray
additive tank in the event of a design-basis accident. Wolf Creek entered this
issue in the corrective action program as Condition Report 38715.
Failure to implement design control measures to analyze whether containment
spray piping remained full of water was a performance deficiency. This finding
was more than minor because it affected the design control attribute of the
Mitigating Systems Cornerstone objective to ensure the availability, reliability,
and capability of the containment spray system to respond to initiating events
and prevent undesirable consequences. Specifically, the inspectors had
reasonable doubt on the capability of the containment spray system to properly
inject because of vortexing in the containment spray additive tank. The
inspectors performed the significance determination using Inspection Manual
Chapter 0609.04. The finding was determined to be of very low safety
significance (Green) because it was a design or qualification deficiency
confirmed not to result in loss of operability or functionality. Although the failure
to have this calculation had existed since original construction, the inspectors
determined this finding reflected current performance since the licensee was
required to evaluate likelihood of tanks allowing gas intrusion into the emergency
core cooling systems in response to Generic Letter 2008-01, Managing Gas
Accumulation in Emergency Core Cooling, Decay Heat Removal, and
Containment Spray Systems. Consequently, this finding had problem
identification and resolution cross-cutting aspects associated with the corrective
action program in that the licensee did not thoroughly evaluate the potential for
gas intrusion from all possible tanks P.1(c)(Section 4OA5).
- 10 - Enclosure
B. Licensee-Identified Violations
Violations of very low safety significance, which were identified by the licensee, have
been reviewed by the inspectors. Corrective actions taken or planned by the licensee
have been entered into the licensees corrective action program. These violations and
condition report numbers are listed in Section 4OA7.
- 11 - Enclosure
REPORT DETAILS
Summary of Plant Status
Wolf Creek began the quarter shut down for Refueling Outage 18. Wolf Creek restarted on
June 22, 2011. Reactor operators manually tripped the reactor from 82 percent power on
June 26 due to a trip of main feedwater pump B. Wolf Creek restarted on June 29 and ended
the quarter holding at 55 percent power to complete troubleshooting and repairs on main
feedwater pump B.
1. REACTOR SAFETY
Cornerstones: Initiating Events, Mitigating Systems, Barrier Integrity, and
1R01 Adverse Weather Protection (71111.01)
.1 Summer Readiness for Offsite and Alternate-ac Power
a. Inspection Scope
The inspectors performed a review of preparations for summer weather for selected
systems, including conditions that could lead to loss-of-offsite power and conditions that
could result from high temperatures. The inspectors reviewed the procedures and
communications protocols between the transmission system operator and the plant to
verify that the appropriate information was being exchanged when issues arose that
could affect the offsite power reliability. Examples of aspects considered in the
inspectors review included:
- The coordination between the transmission system operator and the control
room personnel during off-normal or emergency events
- The explanations for the events
- The estimates of when the offsite power system would be returned to a normal
state
- The notifications from the transmission system operator to the plant when the
offsite power system was returned to normal
During the inspection, the inspectors focused on plant-specific design features and the
procedures used by plant personnel to mitigate or respond to adverse weather
conditions. Additionally, the inspectors reviewed the Updated Safety Analysis
Report (USAR) and performance requirements for selected systems, and verified that
operator actions were appropriate per station procedures. Specific documents reviewed
during this inspection are listed in the attachment. The inspectors also reviewed
corrective action documents to verify that the licensee was identifying adverse weather
issues at an appropriate threshold and entering them into their corrective action
- 12 - Enclosure
program. These activities constitute completion of one readiness for summer weather
affect on offsite and alternate ac power sample as defined in Inspection
Procedure 71111.01-05.
b. Findings
No findings were identified.
.2 Readiness for Impending Adverse Weather Conditions
a. Inspection Scope
When thunderstorms, tornados, and high winds were forecast in the site vicinity on
June 1, 9, and 16, 2011, the inspectors reviewed the plant preparations for the expected
weather conditions. On June 1, 9, and 16, the inspectors walked down the offsite power
system, refueling water storage tank, and reactor makeup water storage tank because
their safety functions could be affected by high wind-generated missiles or a loss of
offsite power. The inspectors evaluated these preparations against the site procedures
and determined that actions by the plant staff were adequate. During the inspection, the
inspectors focused on plant-specific design features and the station procedures used to
respond to specified adverse weather conditions. The inspectors also toured outdoor
areas of the plant to look for any loose debris that could become a wind-driven projectile.
The inspectors evaluated operator staffing and accessibility of instrumentation and
controls for systems required to operate the plant. Additionally, the inspectors reviewed
the USAR and performance requirements for the selected systems and verified that
operator actions were appropriate per station procedures. The inspectors also reviewed
a sample of corrective action documents to verify that the licensee-identified adverse
weather issues at an appropriate threshold and entered them into the corrective action
program. Specific documents reviewed during this inspection are listed in the
attachment.
These activities constitute completion of two readiness for impending adverse weather
condition samples as defined in Inspection Procedure 71111.01-05.
b. Findings
Introduction. The inspectors identified a Green noncited violation of Technical
Specification 5.4.1.a for having no procedure to address onsite debris impacting plant
equipment during severe weather.
Description. On June 1, 2011, a severe thunderstorm watch was announced by the
national weather service. The inspectors walked down the transformer yard at 6 p.m.,
with the storms forecast to arrive later that night. The inspectors found numerous pieces
of unsecured plywood and 2x4 and 2x8 planks. The inspectors brought this to the
licensees attention, and Wolf Creek personnel secured the materials. The inspectors
reviewed station Procedure AI 14-006, Severe Weather, Revision 9A. The procedure
directed public address system announcements for national weather service severe
weather declarations and instructions on personnel sheltering, but included no steps on
- 13 - Enclosure
equipment protection from onsite debris. The inspectors reviewed
Procedure OFN SG-003, Natural Events, Revision 20A, but it did not direct entry until a
tornado is sighted or a tornado warning is issued.
The national weather service issued a tornado warning for the site area on June 9, at
3:20 p.m. The inspectors walked down the transformer yard at 5 p.m. The inspectors
again found unsecured debris in the transformer and tank areas. The inspectors
reported the debris to the control room and outage control center who sent personnel to
secure the material. On June 10, a severe thunderstorm watch was issued at 5 p.m.,
and the inspectors walked down the transformer and tank yards at 6 p.m. to verify the
corrective action from the previous day had been implemented for the pending storms.
The inspectors found that some material was removed or secured, but also found
numerous unsecured sections of scaffolding, wood, palettes, diamond plate, and debris.
The inspectors discussed this with the outage control center. Condition Report 40351
was written with immediate actions to secure the loose materials. The extent of
condition description included any area where inclement weather has the potential of
creating airborne objects that could challenge plant equipment. On June 16, the
inspectors walked down the transformer yard and tank areas during a thunderstorm.
The inspectors found numerous unsecured pieces and stacks of wood and other debris
that could impact plant equipment if winds were more severe. Wolf Creek responded by
securing or removing the debris and writing Condition Report 40573 which implemented
a weekly preventive maintenance activity to clean up outside areas and changed
Procedure AI 14-006 to perform walkdowns of outside areas prior to severe weather.
The inspectors found previously written condition reports on lack of adverse weather
preparations for outdoor areas prior to the inspection.
Analysis. Failure to remove potential wind-driven debris from the transformer and tank
areas before severe weather is a performance deficiency. This finding was more than
minor because it impacted the protection against external factors attribute of the
Initiating Events Cornerstone, and it 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. The inspectors evaluated this finding
using Inspection Manual Chapter 0609.04, and determined that it was of very low safety
significance (Green) for June 16 because it did not contribute to both the likelihood of a
reactor trip and the likelihood that mitigation equipment would be unavailable since the
reactor was shutdown. Inspectors used Manual Chapter 0609, Appendix G, Checklist 4,
for the other occurrences because Wolf Creek was in Modes 4 or 5. The finding again
screened to Green because it did not increase the likelihood of a loss of inventory, did
not cause the loss of reactor coolant system instrumentation, did not degrade the ability
of the licensee to terminate a leak path or add inventory when needed, or degrade the
ability to recover residual heat removal if it was lost. This finding has a cross-cutting
aspect in the area of problem identification and resolution, specifically the corrective
action program attribute because licensee short-term corrective actions failed to ensure
debris was secured or removed prior to severe weather P.1(d).
Enforcement. Technical Specification 5.4.1.a requires, in part, that written procedures
shall be established, implemented, and maintained covering the procedures
recommended in Regulatory Guide 1.33, Revision 2, Appendix A, 1978. Regulatory
- 14 - Enclosure
Guide 1.33, Appendix A, Section 6.w, requires, in part, written procedures for acts of
nature (e.g., tornado, flood, dam failure, earthquakes). Contrary to the above, prior to
June 16, 2011, Wolf Creek had not established written procedures for acts of nature
associated with tornados. Specifically, there were no procedural directions that
addressed how the licensee was to protect from wind-driven projectiles, associated with
tornados, in the protected area. Because this violation was of very low safety
significance and was entered into the licensee's corrective action program as Condition
Report 40573, this violation is being treated as a noncited violation, consistent with
Section 2.3.2 of the NRC Enforcement Policy: NCV 05000485/2011003-01, No
Procedure for Debris in Transformer and Tank Yards Prior to Severe Weather.
1R04 Equipment Alignments (71111.04)
.1 Partial Walkdown
a. Inspection Scope
The inspectors performed a partial system walkdown of the following risk-significant
system:
- March 8, 2011, Component cooling water
The inspectors selected this system based on its risk significance relative to the Reactor
Safety Cornerstone at the time it was 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, USAR, 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 system 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 one partial system walkdown sample as defined
in Inspection Procedure 71111.04-05.
b. Findings
Introduction. The inspectors reviewed a self-revealing Green noncited violation of
Technical Specification 5.4.1a, Administrative Procedures, for an inadequate clearance
order verification which caused a loss of component cooling water B inventory.
- 15 - Enclosure
Description. On March 8, 2011, Wolf Creek was preparing to implement Temporary
Modification Order (TMO) 10-017-EG-00 to install temporary equipment to cool the
radwaste system heat loads. These preparations included hanging clearance
order D-HB-N-029 which required station operators to verify closed manual
valves EGV0079 and HBV0110 and open and uncap the associated piping header hose
connection valves HBV0088 and HBV0111. Until TMO 10-017 is implemented,
component cooling water must be periodically aligned to the radwaste building to cool its
associated nonsafety-related heat loads. This nonsafety component cooling water
function adds seismic vulnerabilities that render the aligned train inoperable
(NCV 05000482/2010007-01). At 9:30 a.m., station operators attempted to move
valve HBV0110 in the closed direction and found that the valve would not turn. The
operators compared the stem position relative to that of an identical model valve. The
operators successfully manipulated travel of valve EGV0079 in the previous step from
the fully open to fully closed position. This apparent position verification was made using
the naked eye, and was the basis for assuming that the valve was firmly on its seat and
signed the clearance order verifications accordingly.
At 2:37 p.m., the control room operators performed a planned routine alignment of
component cooling water train B to radwaste. This action immediately resulted in a
200 gpm component cooling water leak through valve HBV0110 and out of the hose
connection piping penetrations. The rapidly decreasing component cooling water B
surge tank level caused an auto start of the demineralized water makeup to the
component cooling water B surge tank and simultaneously sent an alarm to the control
room operators. However, the demineralized water makeup capacity is only 60 gpm,
resulting in a component cooling water B inventory loss of 140 gpm and a decreasing
surge volume. Without prompt manual actions, the 5000 gallon component cooling
water train B surge tank volume would have been exhausted in 25 minutes, at which
point component cooling water train B would void and fail. For the duration of the leak,
component cooling water train B was unavailable because it was unable to meet its
accident mission time. Operators identified the cause and isolated the component
cooling water supply from the radwaste building. The leak was determined to be
approximately 500 gallons total, or 10 percent of the normal component cooling water
surge tank inventory.
The leak revealed that valve HBV0110 was not fully closed but was stuck in a throttled
position. Station operators were directed by the control room to attempt to move the
valve in the open direction, which they did with an approved torque assist device. When
the operators subsequently moved the valve in the closed direction, it moved beyond its
previous position and was properly seated. Later that evening, when component cooling
water was once again aligned to radwaste, no leakage occurred. Wolf Creek entered
the event into their corrective action as Condition Report 34505.
The inspectors reviewed the history for valve HBV0110. All four of the subject valves
had minimal manipulation since the waste evaporator package they were originally
associated with had been abandoned in place in the early 1990s. Also, periodic
maintenance to inspect and lubricate the valve internals has not been performed during
this time. The last position verification made on valve HBV0110 was conducted April 21,
2006, and indicated that the valve was throttled partially open. The valve was also listed
- 16 - Enclosure
on drawing M-12HB02 as normally throttled. The clearance order paperwork specified
to leave the valve 1.4 turns open upon removal of the clearance order.
The inspectors determined that the operators failed to meet the requirements of station
Procedure AP 21E-001, step 6.4.2.1, to properly position the equipment/components in
the sequence specified on the clearance order tag hang list, as well as step 6.4.3.1, the
independent verification of that component or equipment condition. The inspectors
interviews with operators and station management indicated that the cause of the
leakage was a lack of information communicated to the operators performing the tagout.
Wolf Creek tagout practices did not provide expected, as-found component positions for
taggers and verifiers in the clearance order tag Hang List nor is this information
communicated during pre-job briefings. Wolf Creek initiated Condition Report 40219
which directed oral communication of the expected initial component positions during
pre-job briefings and on the clearance order paperwork template.
Analysis. Failure to properly establish clearance order boundary isolation is a
performance deficiency. The performance deficiency is more than minor because it
impacted the equipment performance and human performance attributes 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 0609.04, the finding was determined
to be of very low safety significance (Green) because the finding is not a design or
qualification deficiency confirmed not to result in loss of operability or functionality; the
finding does not represent a loss of system safety function; the finding does not
represent actual loss of safety function of a single train for more than its technical
specification allowed outage time; the finding does not represent an actual loss of safety
function of one or more nontechnical specification trains of equipment designated as risk
significant per 10 CFR 50.65 for more than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />; and the finding does not screen as
potentially risk significant due to a seismic, flooding, or severe weather initiating event.
The inspectors found that the finding had a cross-cutting aspect of work practices in the
area of human performance. The licensee communicates human error prevention
techniques, such as holding pre-job briefings, self- and peer-checking, and proper
documentation of activities. Specifically, Wolf Creek did not communicate the expected
as-found condition of valve HBV0110 to the taggers and verifiers of the clearance order
Enforcement. Wolf Creek Technical Specification 5.4.1a requires that procedures be
established, implemented, and maintained covering the activities described in
Regulatory Guide 1.33, Revision 2, Appendix A. Regulatory Guide 1.33, Revision 2,
Appendix A, Section 1(c) requires, in part, procedures governing equipment control,
including locking and tagging. Licensee Procedure AP 21E-001 Clearance Orders,
steps 6.4.2.1 and 6.4.3.1, specifies that equipment and components be positioned and
verified in the sequence specified on the clearance order tag list. Contrary to the above,
on March 8, 2011, the licensee failed to ensure the component was positioned and
verified in the sequence specified on the clearance order tag list. Specifically, while
performing clearance order D-HB-N-029, valve HBV0110 was not properly positioned
and verified as specified on the clearance order tag list. These actions directly resulted
in a loss of component cooling water train availability. Because this finding is of very low
- 17 - Enclosure
safety significance and was entered into the licensee corrective action program as
Condition Reports 34505 and 40219, this violation is being treated as a noncited
violation in accordance with Section 2.3.2 of the Enforcement Policy:
NCV 05000482/2011003-02, Failure to Properly Establish Clearance Order Boundary
Isolation Resulting in Loss of Component Cooling Water Inventory.
.2 Complete Walkdown and System Walkdown Associated with Temporary
Instruction (TI) 2515/177
a. Inspection Scope
On April 27, 2011, the inspectors performed a complete system alignment inspection of
the containment spray 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.
The inspectors conducted a walkdown of the containment spray system in sufficient
detail to reasonably assure the acceptability of the licensees walkdowns (TI 2515/177,
Section 04.02.d). The inspectors also verified that the information obtained during the
licensees walkdown was consistent with the items identified during the inspectors
independent walkdown (TI 2515/177, Section 04.02.c.3).
In addition, the inspectors verified that the licensee had isometric drawings that describe
the containment spray system configurations and had acceptably confirmed the
accuracy of the drawings (TI 2515/177, Section 04.02.a). The inspectors verified the
following related to the isometric drawings.
- High point vents were identified
- Other areas where gas can accumulate and potentially impact subject system
operability, such as at orifices in horizontal pipes, isolated branch lines, heat
exchangers, improperly sloped piping, and under closed valves were acceptably
referenced in documentation
- Horizontal pipe centerline elevation deviations and pipe slopes in nominally
horizontal lines that exceed specified criteria were identified
- All pipes and fittings were clearly shown
- 18 - Enclosure
- The drawings were up-to-date with respect to recent hardware changes and that
any discrepancies between as-built configurations and the drawings were
documented and entered into the corrective action program for resolution
The inspectors verified that piping and instrumentation diagrams accurately described
the subject systems; that they were up-to-date with respect to recent hardware changes;
and any discrepancies between as-built configurations, the isometric drawings, and the
piping and instrumentation diagrams were documented and entered into the corrective
action program for resolution (TI 2515/177, Section 04.02.b).
Documents reviewed are listed in the attachment to this report.
These activities constitute completion of one complete system walkdown sample as
defined in Inspection Procedure 71111.04-05. Also, this inspection effort counts toward
the completion of TI 2515/177. See Section 4OA5 for additional information.
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:
- March 19, 2011, Safety injection pump room A
- March 19, 2011, Control room ventilation equipment room B
- March 20, 2011, Auxiliary building 1988 pipe chase
- April 6, 2011, Containment building
The inspectors reviewed these 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
- 19 - Enclosure
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 four quarterly fire-protection inspection samples
as defined in Inspection Procedure 71111.05-05.
b. Findings
No findings were identified.
1R08 Inservice Inspection Activities (71111.08)
.1 Inspection Activities Other Than Steam Generator Tube Inspection, Pressurized Water
Reactor Vessel Upper Head Penetration Inspections, Boric Acid Corrosion Control
(71111.08-02.01)
a. Inspection Scope
The inspection procedure required review of two or three types of nondestructive
examination activities and, if performed, one to three welds on the reactor coolant
system pressure boundary. It also required review of one or two examinations with
relevant indications (if any were found) that had been accepted by the licensee for
continued service.
The inspectors directly observed the following nondestructive examinations:
SYSTEM WELD IDENTIFICATION EXAMINATION TYPE
Pressurizer TBB03-CIRCUM-1-W Ultrasonic Examination
Pressurizer TBB03-SEAM-4W Ultrasonic Examination
Pressurizer TBB03-10-B-W Ultrasonic Examination
Pressurizer TBB03-10-C-W Ultrasonic Examination
Pressurizer TBB03-10-B-IR Ultrasonic Examination
Pressurizer TBB03-10-C-IR Ultrasonic Examination
Steam Generator EBB01A-SEAM-5-W Ultrasonic Examination
- 20 - Enclosure
SYSTEM WELD IDENTIFICATION EXAMINATION TYPE
Steam Generator EBB01A-SEAM-8-W Ultrasonic Examination
RV Closure Head CH-STUD-19 through 36 Ultrasonic Examination
Studs and Nuts
Main Steam AB-01-R001 Visual Examination 3
Piping Support
Main Steam AB-01-R003 Visual Examination 3
Piping Support
Feedwater Piping AE05-R028 Visual Examination 3
Support
Feedwater Piping AE-04-R019 Visual Examination 3
Support
Feedwater Piping AE05-C001 Visual Examination 3
Support
Main Steam AB-01-R010 Magnetic Examination
Integral
Attachment
Feedwater AE-05-R028 Magnetic Examination
Integral
Attachment
During the review and observation of each examination, the inspectors verified that
activities were performed in accordance with ASME Boiler and Pressure Vessel Code
requirements and applicable procedures. Indications were compared with previous
examinations and dispositioned in accordance with ASME code and approved
procedures. The qualifications of all nondestructive examination technicians performing
the inspections were verified to be current.
Only the visual examination of AE05-R028, Piping Support, identified any relevant
indications. Repairs were made to AE05-R028 and it was reexamined and was
satisfactory. Wolf Creek personnel stated that no relevant indications were accepted by
the licensee for continued service.
- 21 - Enclosure
The inspectors directly observed a portion of the following welding activities:
SYSTEM WELD IDENTIFICATION WELD TYPE
Reactor Coolant 13BG22-W-33 Shield Metal Arc Welding
System
Reactor Coolant 13BG22-W-34 Shield Metal Arc Welding
System
The inspectors verified, by review, that the welding procedure specifications and the
welders had been properly qualified in accordance with ASME Code,Section IX,
requirements. The inspectors also verified through 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.
b. Findings
.1 Failure to Ensure Fillet Weld Met Size Requirements on Train B Charging Header Vent
Line
Introduction. The inspectors documented a self-revealing Green noncited violation of
10 CFR Part 50, Appendix B, Criterion IX, Control of Special Processes, after the
licensee failed to ensure that that the fillet weld between the train B charging header and
the half-coupling used to attach two vent valves met 2:1 taper weld requirements. The
undersized weld subsequently resulted in a 300 drop-per-minute leak in January 2011.
Description. On January 3, 2011, the licensee identified a 300 drop-per-minute pinhole
leak at the weld joint between the train B charging header and/or the half coupling used
to attach vent valves BGV0845 and BGV0846. The licensee measured the subject weld
and concluded that the weld was undersized and the required 2:1 aspect ratio was not
obtained. The weld was performed in the October/November 2009 timeframe during the
installation of vent valves in the chemical and volume control system, the residual heat
removal system, and the high pressure coolant injection system. Also, quality assurance
inspectors inappropriately accepted the undersized weld.
Wolf Creeks extent-of-condition review concluded that 12 additional welds either did not
meet the procedurally required 2:1 aspect ratio or did not meet ASME minimum weld
size requirements. No other undersized welds developed leaks. After the leak was
identified, the train B charging system was declared inoperable and the weld was
repaired and built up to the correct 2:1 aspect ratio.
Wolf Creek performed a hardware failure analysis on the failed weld and concluded that
although the main characteristics of the weld fracture were consistent with stress
corrosion cracking, the crescent shape of the fracture indicated cyclic crack growth. The
licensee also concluded that the configuration of the vent line with no lateral support
- 22 - Enclosure
could have created a cantilever effect on the line and in combination with a notch
created by the lack of fusion in the weld root served as a stress concentrator. This issue
was entered into the licensees corrective action program as Condition Report 36438.
Analysis. Failure to meet ASME code weld size requirements is a performance
deficiency. The finding was more than minor because it was associated with the
equipment performance attribute of the Initiating Events Cornerstone. The finding
adversely affected the cornerstone objective to limit the likelihood of those events that
upset plant stability and challenge critical safety functions during power operations. The
inspectors performed a Phase 1 screening in accordance with Inspection Manual
Chapter 0609.04 and determined that the finding was of very low safety significance
(Green) because the issue did not result in exceeding the technical specification limit for
identified reactor coolant system leakage or affect other mitigating systems resulting in a
total loss of their safety function. This finding had a cross-cutting resources aspect in
the area of human performance, because the licensee failed to ensure that welders and
quality assurance inspectors were adequately trained in the procedural requirements
and methods for measuring weld dimensions to assure nuclear safety H.2(b).
Enforcement. Title 10 CFR Part 50, Appendix B, Criterion IX, Control of Special
Processes, requires in part, that measures be established to ensure that special
processes, including welding are controlled and accomplished by qualified personnel
using qualified procedures in accordance with applicable codes, standards,
specifications, criteria, and other special requirements. Contrary to the above, in
October 2009, the licensee failed to ensure that special processes, including welding,
were controlled and accomplished using qualified procedures. Specifically, welders
failed to ensure that the fillet weld between the train B charging header and the half-
coupling used to attach two vent valves met 2:1 taper weld requirements, which
subsequently resulted in a 300 drop-per-minute leak in January 2011. This issue was
entered into the licensees corrective action program as Condition Report 36438.
Because this finding was determined to be of very low safety significance and was
entered into the licenses corrective action program, this violation is being treated as a
noncited violation consistent with Section 2.3.2 of the NRC Enforcement Policy:
NCV 05000482/2011003-03, Failure to Assure Fillet Weld Met Size Requirements on
Train B Charging Header Vent Line.
.2 Failure to Ensure Separation of Stainless Steel and Carbon Steel Grinding and Cutting
Tools
Introduction. The inspectors identified a Green noncited violation of 10 CFR 50.55a,
Codes and Standards, after the licensee failed to ensure that stainless steel and
carbon steel grinding wheels, flapper wheels, cutting wheels, and files were stored
separately and used only for the weld preparation of the designated steel.
Description. During inspection of the tool issue room in the radiologically controlled
area, the inspectors identified that tools designated for either stainless steel or carbon
steel weld preparation were not stored separately. The inspectors noted that although
stainless steel grinding wheels, flapper wheels, and cutting wheels were marked for
stainless steel use, they were stored with carbon steel grinding wheels, flapper wheels,
- 23 - Enclosure
and cutting wheels. Additionally, the inspectors identified that although stainless steel
files and carbon steel files were stored in separate drawers, there were files in the
stainless steel drawer that appeared to have been used on carbon steel, and there was
a file marked for use on stainless steel in the carbon steel drawer. There was also no
procedure or written guidance pertaining to proper storage and control of the equipment.
The failure to separate tools used for stainless steel weld preparation from tools used for
carbon steel preparation could result in the contamination of stainless steel welds by
carbon steel and affect the material integrity and corrosion resistance. The licensee
immediately removed the tools and replaced them with new tools stored separately for
use on specific types of metals. This issue was entered into the licensees corrective
action program as Condition Report 3644.
Analysis. Failure to protect stainless steel welds from deleterious contamination is a
performance deficiency. The finding was more than minor because it was associated
with the equipment performance attribute of the Initiating Events Cornerstone. The
finding adversely affected the cornerstone objective to limit the likelihood of those events
that upset plant stability and challenge critical safety functions during power operations
and if left uncorrected, the finding would become a more significant safety concern. The
inspectors performed a Phase 1 screening in accordance with Inspection Manual
Chapter 0609.04 and determined that the finding was of very low safety significance
(Green) because the issue did not result in exceeding the technical specification limit for
identified reactor coolant system leakage or affect other mitigating systems resulting in a
total loss of their safety function. This finding had a resources cross-cutting aspect in
the area of human performance, because the licensee did not provide adequate
procedures for the preparation of stainless steel and carbon steel welds H.2(c).
Enforcement. Title 10 CFR 50.55a, states in part, that Each operating license for a
boiling or pressurized water-cooled nuclear power facility is subject to the conditions in
paragraphs (f) and (g). Title 10 CFR 50.55a(g)(4), requires, in part, that components
classified as ASME Code Class 1, Class 2, and Class 3 meet the requirements set forth
in Section XI of the ASME Boiler and Pressure Vessel Code and Addenda.Section XI of
the ASME Code, Part IWA-4221(b)(2), states that When adding a new component to an
existing system, the Owner shall specify a Construction Code. The licensee specified
Section III of the subject code when adding a new component to an existing system.
Section III, Part NG4412, states that The work [weld preparation] shall be protected
from deleterious contamination. Contrary to the above, prior to June 2011, the licensee
did not ensure that weld preparation was protected from deleterious contamination.
Specifically, the licensee failed to ensure weld preparation was protected, in that tools
located in the hot tool room drawers containing files, grinding wheels, flapper wheels,
and cutting wheels that were used for the purpose of weld preparation, were found to
contain a mixture of both stainless steel tools and carbon steel tools. This issue was
entered into the licensees corrective action program as Condition Report 36444.
Because this finding was determined to be of very low safety significance and was
entered into the licensees corrective action program, this violation is being treated as a
noncited violation consistent with Section 2.3.2 of the NRC Enforcement Policy:
NCV 05000482/2011003-04, Failure to Assure Separation of Stainless Steel and
Carbon Steel Grinding and Cutting Equipment.
- 24 - Enclosure
.2 Vessel Upper Head Penetration Inspection Activities (71111.08-02.02)
a. Inspection Scope
The inspectors witnessed portions of the licensees performance of the required visual
inspection (VT-2) of the reactor head and pressure-retaining components above the
reactor pressure vessel head in accordance with requirement of ASME Code
Case N-729-1 as mandated by 10 CFR 50.55a. Implementation required ASME
Code IWA-2212 VT-2 under the mirror insulation on top of the reactor head through
multiple access points. The inspectors reviewed the results of this inspection for
evidence of leaks or boron deposits at reactor pressure boundaries and related
insulation above the head. Specific documents reviewed during this inspection are listed
in the attachment.
These actions constitute completion of the requirements for Section 02.02 of Inspection
b. Findings
No findings were identified.
.3 Boric Acid Corrosion Control Inspection Activities (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 reviewed the documentation associated
with the licensees boric acid corrosion control walkdown as specified in
Procedure STN PE-040D, RCS Pressure Boundary Integrity Walkdown, Revision 3,
and Procedure AP 16F-001, Boric Acid Corrosion Control Program, Revision 6A. 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.
These actions constitute completion of the requirements for Section 71111.08-02.03.
b. Findings
Failure to Assure Configuration Control of Safety-Related Systems
Introduction. The inspectors identified a Green noncited violation of 10 CFR Part 50,
Appendix B, Criterion III, Design Control, involving the failure of the licensee to review
- 25 - Enclosure
the suitability of replacing the design specified stainless steel manifold plugs with test
fittings and brass caps on various flow transmitter equalizing block valve drain ports.
Description. During a boric acid walkdown, the inspectors identified that drain ports on
the equalizing block of two separate reactor coolant system flow transmitters had brass
fittings installed instead of stainless steel fittings. The inspectors brought this condition
to Wolf Creeks attention. The licensee determined that a design configuration
nonconformance existed in that licensee Drawing J-17D22 specified that stainless steel
manifold plugs be installed in the drain ports during plant operation. The licensee failed
to review the suitability of installing brass fittings and leaving test fittings on flow
transmitter equalizing block valve drain ports instead of the design specified stainless
steel manifold plugs. Wolf Creek immediately replaced the brass caps with stainless
steel fittings. This issue was entered into the licensees corrective action program as
Condition Report 36439.
Analysis. Plugging instrument lines with test fittings of a different material is a
performance deficiency. The finding was more than minor because it was associated
with the design control attribute of the Initiating Events Cornerstone. The finding
affected the cornerstone objective to limit the likelihood of those events that upset plant
stability and challenge critical safety functions during power operations. The inspectors
screened the finding per Inspection Manual Chapter 0609.04 and determined that the
finding was of very low safety significance (Green) because the issue would not result in
exceeding the technical specification limit for identified reactor coolant system leakage
or affect other mitigating systems resulting in a total loss of their safety function. The
inspectors also determined that the finding had a resources cross-cutting aspect in the
area of human performance, because the licensee did not provide adequate training of
personnel so that the inappropriately installed fittings could be identified during system
walkdowns H.2(b).
Enforcement. Title 10 CFR Part 50, Appendix B, Criterion III, Design Control, requires,
in part, that measures be established for the selection and review of suitability of
application of materials, parts, equipment, and processes that are essential to the safety-
related functions of the structures, systems, and components. Contrary to the above,
the licensee failed to establish measures for the selection and review for suitability of
parts that are essential to the safety-related functions of systems and components.
Specifically, the licensee failed to review the suitability of replacing the design specified
stainless steel manifold plugs with brass caps and test fittings on various equalizing
block valve drain ports for pressure, differential pressure, and flow transmitters. This
issue was entered into the licensees corrective action program as Condition
Report 36439. Because this finding was determined to be of very low safety significance
and was entered into the licensees corrective action program, this violation is being
treated as a noncited violation consistent with Section 2.3.2 of the NRC Enforcement
Policy: NCV 05000482/2011003-05, Failure to Assure Configuration Control of Safety-
Related Systems.
- 26 - Enclosure
.4 Steam Generator Tube Inspection Activities (71111.08-02.04)
a. Inspection Scope
The inspection procedure specified an assessment of in situ screening criteria to assure
consistency between assumed nondestructive examination flaw sizing accuracy and
data from the EPRI examination technique specification sheets. The inspection
procedure also specified assessment of appropriateness of tubes selected for in situ
pressure testing, observation of in situ pressure testing, and review of in situ pressure
test results. No conditions were identified that warranted in situ pressure testing. The
steam generators are original construction steam generators.
The inspectors reviewed both the licensee site-validated and qualified acquisition and
analysis technique sheets used during this refueling outage and the qualifying EPRI
examination technique specification sheets to verify that the essential variables
regarding flaw sizing accuracy, tubing, equipment, technique, and analysis had been
identified and qualified through demonstration.
Wolf Creek completed steam generator eddy current inspections for Refueling
Outage 18 on April 12, 2011. In accordance with the EPRI steam generator examination
guidelines, bobbin coil inspections were expanded in steam generator B due to
inspection results associated with wear at anti-vibration bar locations that resulted in a
C-2 condition. In accordance with the EPRI guidelines, another 20 percent of the tubing
in steam generator B was inspected. No other scope expansions were required. In
accordance with Technical Specification 5.5.9.c, nine tubes in steam generator B, three
tubes in steam generator C, and three tubes in steam generator D were plugged based
on inspection results indicating they contained flaws with a depth equal to or exceeding
40 percent of the nominal tube wall thickness. The damage mechanism associated with
each of the pluggable indications was wear at anti-vibration bar locations. No tubes in
steam generator A required plugging. No new corrosion damage mechanisms were
identified.
The following secondary side maintenance and inspections were performed:
- Sludge lancing of all four steam generators. The amount of sludge removed from
each steam generator was: steam generator A, 26 lbs; steam generator B,
34 lbs; steam generator C, 30 lbs; and steam generator D, 27.5 lbs.
- Foreign object search and retrieval of all four steam generators to locate, identify,
and retrieve foreign objects present on the steam generator tube sheet. Foreign
object search and retrieval was also performed to inspect for any possible loose
parts identified during the eddy current program. Minor foreign objects were
identified and addressed within the corrective action program and plant
procedures. Visual examination and eddy current testing verified that no
degradation was associated with any tubes surrounding the foreign objects.
- In-bundle inspection of steam generators B and C to inspect the condition of the
top of the tube sheet and to augment the foreign object search and retrieval
- 27 - Enclosure
effort. No anomalies were identified during these inspections and the information
will be used for trending and to plan future maintenance operations.
- Upper steam drum inspection in steam generators B and C to evaluate the
condition of the upper steam drum components with regard to damage of any
kind. Ultrasonic testing was also performed on locations susceptible to erosion
on the feeding in steam generators B and C. No anomalies were identified and
the information will be used for trending and to plan future maintenance.
Specific documents reviewed during this inspection are listed in the attachment.
These actions constitute completion of the requirements for Section 71111.08-02.04.
b. Findings
No findings were identified.
.5 Identification and Resolution of Problems (71111.08-02.05)
a. Inspection Scope
The inspectors reviewed 99 condition reports which dealt with inservice inspection
activities and found the corrective actions for inservice inspection issues were
appropriate. The specific condition reports reviewed are listed in the documents
reviewed section. From this review the inspectors concluded that the licensee had an
appropriate threshold for entering inservice inspection issues into the corrective action
program and had procedures that direct a root cause evaluation when necessary. The
licensee also had an effective program for applying industry inservice inspection
operating experience. Specific documents reviewed during this inspection are listed in
the attachment.
These actions constitute completion of the requirements for Section 71111.08-02.05.
b. Findings
No findings were identified.
1R11 Licensed Operator Requalification Program (71111.11)
.1 Inspection Scope
On June 14, 2011, the inspectors observed a crew of licensed operators in the plants
simulator to verify that operator performance was adequate, evaluators were identifying
and documenting crew performance problems; and training was being conducted in
accordance with licensee procedures. The inspectors evaluated the following areas:
- Licensed operator performance
- Crews clarity and formality of communications
- 28 - Enclosure
- Crews ability to take timely actions in the conservative direction
- Crews prioritization, interpretation, and verification of annunciator alarms
- Crews correct use and implementation of abnormal and emergency procedures
- Control board manipulations
- Oversight and direction from supervisors
- Crews ability to identify and implement appropriate technical specification
actions and emergency plan actions and notifications
- Compliance with assumptions for manual action timing in Chapter 15 of the
The inspectors compared the crews performance in these areas to preestablished
operator action expectations and successful critical task completion requirements.
Specific documents reviewed during this inspection are listed in the attachment.
These activities constitute completion of one quarterly licensed-operator requalification
program 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 system:
- Vital switchgear air conditioning units
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:
- Implementing appropriate work practices
- Identifying and addressing common cause failures
- Scoping of systems in accordance with 10 CFR 50.65(b)
- 29 - Enclosure
- Characterizing system reliability issues for performance
- Charging unavailability for performance
- 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 one 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:
- June 3, 2011, Component cooling water train A while train B was inoperable
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
- 30 - Enclosure
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 one maintenance risk assessment and
emergent work control inspection sample as defined in Inspection
Procedure 71111.13 05.
b. Findings
No findings were identified.
1R15 Operability Evaluations (71111.15)
a. Inspection Scope
The inspectors reviewed the following issues:
- March 12, 2011, Emergency diesel generator A jacket water leak
- May 18, 2011, Source range NI-31 high counts after loss of cavity cooling
- June 16, 2011, Essential service water system flaw evaluations
The inspectors selected these potential operability issues based on the risk significance
of the associated components and systems. The inspectors evaluated the technical
adequacy of the evaluations to ensure that technical specification operability was
properly justified and the subject component or system remained available so that no
unrecognized increase in risk occurred. The inspectors compared the operability and
design criteria in the appropriate sections of the technical specifications and USAR to
the licensee personnels evaluations to determine whether 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. The inspectors determined, where appropriate,
compliance with bounding limitations associated with the evaluations. Additionally, the
inspectors also 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 71111.15-04.
b. Findings
No findings were identified.
- 31 - Enclosure
1R18 Plant Modifications (71111.18)
a. Inspection Scope
To verify that the safety functions of important safety systems were not degraded, the
inspectors reviewed the temporary modification identified as TMO 10-017, component
cooling water modification to radioactive waste building.
The inspectors reviewed the temporary modification and the associated safety-
evaluation screening against the system design bases documentation, including the
USAR and the technical specifications, and verified that the modification did not
adversely affect the system operability/availability. The inspectors also verified that the
installation and restoration were consistent with the modification documents and that
configuration control was adequate. Additionally, the inspectors verified that the
temporary modification was identified on control room drawings, appropriate tags were
placed on the affected equipment, and licensee personnel evaluated the combined
effects on mitigating systems and the integrity of radiological barriers.
These activities constitute completion of one sample for temporary plant modifications as
defined in Inspection Procedure 71111.18-05.
b. Findings
No findings were identified.
.2 Permanent Modifications
a. Inspection Scope
The inspectors reviewed key parameters associated with energy needs, materials,
timing, heat removal, control signals, licensing basis, and failure modes for the
permanent modification identified as the source range gamma metrics equivalency to
Westinghouse detectors.
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; systems, structures and components
performance characteristics still meet the design basis; the modification design
assumptions were appropriate; 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 one sample for permanent plant modifications
as defined in Inspection Procedure 71111.18-05.
- 32 - Enclosure
b. Findings
No findings were identified.
1R19 Postmaintenance Testing (71111.19)
a. Inspection Scope
The inspectors reviewed the following postmaintenance activities to verify that
procedures and test activities were adequate to ensure system operability and functional
capability:
- November 17, 2010, Startup feedwater pump testing after rebuild
- March 7, 2011, Feedwater regulating bypass valve controller setting adjustment
- April 1, 2011, Solid state protection system train B after Westinghouse card
testing
- May 17, 2011, Offsite power to engineered safety features transformer A after
replacement of Raychem splices
- June 12, 2011, Component cooling water to thermal barrier heat exchangers
after flow balance Procedure SYS EG-205
- June 24, 2011, Main turbine overspeed testing after replacement
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:
- 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 USAR,
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 postmaintenance 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.
- 33 - Enclosure
These activities constitute completion of six postmaintenance testing inspection samples
as defined in Inspection Procedure 71111.19-05.
b. Findings
Introduction. The inspectors identified a finding of very low safety significance (Green)
involving the failure to follow the requirements of Procedure AP 16E-002, Post
Maintenance Testing Development, for the startup feedwater pump.
Description. On November 4-6, 2010, Wolf Creek workers performed maintenance on
the startup feedwater pump to replace a leaking pump casing gasket. Workers
disassembled the pump per the vendor manual instructions and found a split casing
gasket and both mechanical seals darkened and cracked from overheating. The pump
was reassembled using new parts including bearings, O-rings, mechanical seals, and
casing gasket. The service water cooling lines were also replaced. Wolf Creek
Procedure AP 16E-002, Post Maintenance Testing Development, states that it provides
guidelines to develop test criteria for plant equipment after the completion of
maintenance activities. The procedure also ensures proper testing is implemented to
prove components, systems, and sub-systems perform as designed after the completion
of maintenance activities. Furthermore, Procedure AP 16E-002, Revision 9C, step 6.2
and attachments, requires that when a pump is disassembled or replaced, the
postmaintenance testing includes a pump-flow capacity test be conducted to determine
the capability of the pump to produce the required flow rates within the range of
differential pressure limits. Also, it requires that vibration, bearing, and lubrication
temperatures, motor current, external leakage, and lubrication level are satisfactory.
The inspectors reviewed root cause corrective action 25817-02-14 which created
Procedure STN AE-007 to test the pump with no local actions and ensure a minimum
recirculation flow of 60,000 pounds per hour for pump protection. The inspectors did not
find a discussion of adequate flow. On November 17, 2010, Wolf Creek conducted
surveillance Procedure STN AE-007, Startup Main Feedwater Pump Operational Test,
following the pump reassembly. The only acceptance criteria listed in this procedure
was that the motor-driven feedwater pump started from the control room with no local
operator action. The test contained no acceptance criteria to ensure that after the
completion of maintenance activities, the pump could produce the required flow rates for
either low power or emergency operations.
The purpose of the motor-driven startup feedwater pump is to provide heated feedwater
to the steam generators during plant startup and shutdown operations. The startup
feedwater pump is a horizontal, multi-stage, centrifugal pump with a rated maximum flow
rate of 260,000 pounds per hour. Maximum flow through the startup feedwater pump
suction lines is limited to 230,000 pounds per hour to prevent excessive tube vibration in
the steam generator blowdown regenerative heat exchanger. According to Wolf Creek
training materials, Form APF 30E-004-01, Revision 2, Main Feedwater System, the
required steam generator flow rate during plant startup is 200,000 pounds per hour.
This is based on the maximum steam generator blowdown rate, the heat lost to ambient
surroundings from all main steam lines, and the maximum heatup rate of all main steam
lines and the turbine stop valve heat. The startup feed pump is also used in Emergency
- 34 - Enclosure
Management Guideline FR-H1, Response to Loss of Secondary Heat Sink, step 17, to
feed the steam generator until the steam generator level is restored to greater than the
minimum level for ensuring an adequate heat sink. The success criteria in emergency
operating procedures for feedwater is based on 270,000 pounds per hour for auxiliary
feedwater or greater than 6 percent narrow range steam generator level. The
emergency operating procedure setpoint document requires 250,000 pounds per hour
from each motor-driven auxiliary feedwater pump.
This issue is captured in Condition Report 39494. Wolf Creek issued a new work
package to conduct a single-point pump capacity test and complete the required
postmaintenance testing in accordance with Procedure AP 16E-002. Wolf Creek also
found that there was not a technical basis for the blowdown heat exchanger vibrations
which previously limited the allowable flow through the pump to approximately
200,000 pounds per hour. Wolf Creek initial calculations, pending final review, show that
the pump would be capable of enough flow to provide a heat sink.
Analysis. The failure to follow Procedure AP 16E-002 for developing test criteria for
plant equipment after the completion of maintenance activities is a performance
deficiency. The finding is more than minor because it is associated with the Mitigating
Systems Cornerstone attribute of equipment performance and it adversely affects the
cornerstone objective of ensuring the availability, reliability, and capability of systems
that respond to initiating events to prevent undesirable consequences. Using Inspection
Manual Chapter 0609.04, the inspectors determined that the finding had very low safety
significance (Green) because it did not result in a loss of system safety function, an
actual loss of safety function of a single train for greater than its technical specification
allowed outage time, or screen as potentially risk significant due to a seismic, flooding,
or severe weather initiating event. The inspectors determined that the finding had a
cross-cutting aspect in the area of problem identification and resolution. Specifically,
Wolf Creek created a testing procedure in response to a root cause evaluation, but did
not consider acceptance criteria to ensure that the pump performs acceptably P.1(d).
Enforcement. Enforcement action does not apply because the performance deficiency
did not involve a violation of regulatory requirements. This finding is of very low safety
significance and the issue was entered into the licensee's corrective action program as
Condition Report 39494: FIN 05000482/2011003-06, Inadequate Acceptance Criteria
for Postmaintenance Testing of the Startup Feedwater Pump.
1R20 Refueling and Other Outage Activities (71111.20)
a. Inspection Scope
The inspectors reviewed the outage safety plan and contingency plans for the refueling
outage, conducted from March 19 through June 22, 2011, and June 26-29, 2011, to
confirm that licensee personnel had appropriately considered risk, industry experience,
and previous site-specific problems. The inspectors determined that the plan ensured
sufficient defense in depth. During the refueling outage, the inspectors observed
portions of the shutdown and cooldown processes and monitored licensee controls over
the outage activities listed below.
- 35 - Enclosure
- Configuration management maintains defense in depth, commensurate with the
outage safety plan, and in compliance with technical specifications.
- Clearance activities were properly tagged and equipment configured to safely
support work.
- 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 power ascension, tracking of startup prerequisites, walkdown of
containment to verify that debris removal, 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 sample and one forced
outage inspection sample as defined in Inspection Procedure 71111.20-05.
b. Findings
No findings were identified.
- 36 - Enclosure
1R22 Surveillance Testing (71111.22)
a. Inspection Scope
The inspectors reviewed the USAR, 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 supported operability or functionality
- Test equipment removal
- Restoration of plant systems
- Fulfillment of ASME code requirements
- Updating of performance indicator data
- Engineering evaluations, root causes, and bases for restoring systems,
structures, and components not meeting acceptance criteria were correct
- Reference setting data
- Annunciator and alarm setpoints
The inspectors also verified that licensee personnel identified and implemented any
needed corrective actions associated with the surveillance testing. The following
surveillance testing was observed:
- September 8, 2010, Main steam valve AB-V85 inservice valve test
- 37 - Enclosure
- April 24, 2011, Residual heat removal room cooler B test
- April 26, 2011, Filling, venting, and void surveillance of safety injection
- May 13, 2011, Tan-delta testing of offsite power underground cables
- May 14, 2011, STS PE-018, Containment integrated leak rate test
- May 18, 2011, Containment isolation valve EJHV8811B inservice test
- May 24, 2011, Filling, venting, and void surveillance of residual heat removal
train B
- June 15, 2011, STS IC-615A, Safety injection signal slave relay test
Specific documents reviewed during this inspection are listed in the attachment.
These activities constitute completion of eight surveillance testing inspection samples as
defined in Inspection Procedure 71111.22-05.
b. Findings
No findings were identified.
.2 Surveillance Testing Associated with TI 2515/177, Managing Gas Accumulation in
Emergency Core Cooling, Decay Heat Removal, and Containment Spray Systems
a. Inspection Scope
When reviewing the April 26, 2011, filling, venting, and void surveillance of safety
injection and the May 24, 2011, filling, venting, and void surveillance of residual heat
removal train B surveillances listed above in Section 1R22.1, the inspectors verified that
the procedures were acceptable for (1) testing with shutdown operation, maintenance,
and subject system modifications; (2) void determination and elimination methods; and
(3) post-event evaluation.
The inspectors reviewed the procedures used for conducting surveillance tests and the
determination of void volumes to ensure that the acceptance criteria were satisfied and
would be reasonably assured to remain satisfied until the next scheduled surveillance test
(TI 2515/177, Section 04.03.a). Also, the inspectors reviewed procedures used for filling
and venting following conditions which may have introduced voids into the subject
systems to verify that the procedures acceptably addressed testing for such voids and
provided acceptable processes for their reduction or elimination (TI 2515/177,
Section 04.03.b). Specifically, the inspectors verified that:
- 38 - Enclosure
- Gas intrusion prevention, refill, venting, monitoring, trending, evaluation, and void
correction activities were acceptably controlled by approved operating
procedures (TI 2515/177, Section 04.03.c.1)
- Procedures ensured the system did not contain voids that may jeopardize
operability (TI 2515/177, Section 04.03.c.2)
- Procedures established that void criteria were satisfied and will be reasonably
ensured to be satisfied until the next scheduled void surveillance (TI 2515/177,
Section 04.03.c.3)
- The licensee entered changes into the corrective action program as needed to
ensure acceptable response to issues. In addition, the inspectors confirmed that
a clear schedule for completion is included for corrective action program entries
that have not been completed (TI 2515/177, Section 04.03.c.5)
- Procedures included independent verification that critical steps were completed
(TI 2515/177, Section 04.03.c.6)
The inspectors verified the following with respect to surveillance and void detection:
- Specified surveillance frequency was consistent with technical specification
requirements (TI 2515/177, Section 04.03.d.1)
- Surveillance frequencies were stated or, when conducted more often than
required by technical specifications, the process for their determination was
described (TI 2515/177, Section 04.03.d.2)
- Surveillance methods were acceptably established to achieve the needed
accuracy (TI 2515/177, Section 04.03.d.3)
- Surveillance procedures included up-to-date acceptance criteria (TI 2515/177,
Section 04.03.d.4)
- Procedures included effective follow-up actions when acceptance criteria are
exceeded or when trending indicates that criteria may be approached before the
next scheduled surveillance (TI 2515/177, Section 04.03.d.5)
- Measured void volume uncertainty was considered when comparing test data to
acceptance criteria (TI 2515/177, Section 04.03.d.6)
- Venting procedures and practices utilized criteria such as adequate venting
durations and observing a steady stream of water (TI 2515/177,
Section 04.03.d.7)
- 39 - Enclosure
- An effective sequencing of void removal steps was followed to ensure that gas
does not move into previously filled system volumes (TI 2515/177,
Section 04.03.d.8)
- Qualitative void assessment methods included expectations that the void will be
significantly less than allowed by acceptance criteria (TI 2515/177,
Section 04.03.d.9)
- Venting results were trended periodically to confirm that the systems are
sufficiently full of water and that the venting frequencies are adequate. The
inspectors also verified that records on the quantity of gas at each location are
maintained and trended as a means of pre-emptively identifying degrading gas
accumulations (TI 2515/177, Section 04.03.d.10)
- Surveillances were conducted at any location where a void may form, including
high points, dead legs, and locations under closed valves in vertical pipes
(TI 2515/177, Section 04.03.d.11)
- The licensee ensured that systems were not preconditioned by other procedures
that may cause a system to be filled, such as by testing, prior to the void
surveillance (TI 2515/177, Section 04.03.d.12)
- Procedures included gas sampling for unexpected void increases if the source of
the void is unknown and sampling is needed to assist in determining the source
(TI 2515/177, Section 04.03.d.13)
The inspectors verified the following with respect to filling and venting:
- Revisions to fill and vent procedures to address new vents or different venting
sequences were acceptably accomplished (TI 2515/177, Section 04.03.e.1)
- Fill and vent procedures provided instructions to modify restoration guidance to
address changes in maintenance work scope or to reflect different boundaries
from those assumed in the procedure (TI 2515/177, Section 04.03.e.2)
The inspectors verified the following with respect to void control:
- Void removal methods were acceptably addressed by approved procedures
(TI 2515/177, Section 04.03.f.1)
- The licensee had reasonably ensured that the residual heat removal pump is free
of damage following a gas-related event in which pump acceptance criteria was
exceeded (TI 2515/177, Section 04.03.f.2)
Documents reviewed are listed in the attachment to this report.
This inspection effort counts towards the completion of TI 2515/177. See Section 4OA5
for additional information.
- 40 - Enclosure
b. Findings
No findings were identified.
Cornerstone: Emergency Preparedness
1EP4 Emergency Action Level and Emergency Plan Changes (71114.04)
a. Inspection Scope
The inspector performed an in-office review of the Wolf Creek APF 06-002-01,
Emergency Action Levels, Revision 15A. This revision made two administrative
changes to EAL-6, Loss of Electrical Power/Assessment Capability. The change
included replacing the abbreviation D/Gs with the capitalized and bolded wording
DIESEL GENERATORS, and capitalizing and bolding the wording NB
TRANSFORMERS.
This revision was compared to its previous revision, to the criteria of NUREG-0654,
Criteria for Preparation and Evaluation of Radiological Emergency Response Plans and
Preparedness in Support of Nuclear Power Plants, Revision 1, and to the standards in
10 CFR 50.47(b) to determine if the revision adequately implemented the requirements
of 10 CFR 50.54(q). This review was not documented in a safety evaluation report and
did not constitute approval of licensee-generated changes; therefore, this revision is
subject to future inspection.
These activities constitute completion of one sample as defined in Inspection
Procedure 71114.04-05.
b. Findings
No findings were identified.
4. OTHER ACTIVITIES
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 1st Quarter 2011 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.
- 41 - Enclosure
b. Findings
No findings were identified.
.1 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 2nd Quarter 2010 through the 1st
Quarter 2011. The inspectors used definitions and guidance contained in NEI
Document 99-02, Regulatory Assessment Performance Indicator Guideline, Revision 6
to determine the accuracy of the performance indicator data reported during those
periods. 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 1, 2010, through March 31, 2011, to
validate the accuracy of the submittals. The inspectors also reviewed the licensees
condition 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 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.
.2 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 2nd Quarter 2010 through the
1st Quarter 2011. The inspectors used definitions and guidance contained in NEI
Document 99-02, Regulatory Assessment Performance Indicator Guideline, Revision 6
to determine the accuracy of the performance indicator data reported during those
periods. 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 1, 2010, through March 31, 2011, to validate the accuracy
of the submittals. The inspectors also reviewed the licensees condition 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.
- 42 - Enclosure
b. Findings
No findings were identified.
4OA2 Identification and Resolution of Problems (71152)
Cornerstones: Initiating Events, Mitigating Systems, Barrier Integrity, Emergency
Preparedness, Public Radiation Safety, Occupational Radiation Safety, and Physical
Protection
.1 Routine Review of Identification and Resolution of Problems
a. Inspection Scope
As part of various baseline inspections discussed in previous sections, the inspectors
reviewed issues to verify that they were being entered into the Wolf Creek corrective
action program at an appropriate threshold. The inspectors verified the program to be
addressing issues in a timely manner as well as identifying and correcting adverse
trends. The inspectors reviewed attributes that included:
- Complete and accurate identification of the problem
- Timely correction, commensurate with the safety significance
- Evaluation and disposition of performance issues, generic implications, common
causes, contributing factors, root causes, extent of condition reviews, and
previous occurrences reviews
- Classification, prioritization, focus, and timeliness of corrective actions.
Minor issues entered into the licensees corrective action program because of the
inspectors observations are included in the attached list of documents reviewed.
These reviews for the identification and resolution of problems did not constitute any
additional inspection samples. They were considered a 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
The inspectors performed a daily screening of items entered into the licensees
corrective action program through review of the Wolf Creeks daily corrective action
documents.
- 43 - Enclosure
The inspectors performed these daily reviews as part of their plant status monitoring
activities and 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 Wolf Creek corrective action program and
associated documents to identify trends that could represent a more significant safety
issue. The inspectors focused their review on repetitive equipment issues, but also
considered the results of daily corrective action item screenings, licensee trending
efforts, and licensee human performance results. The inspectors considered the
6-month period of January through June 2011 although some examples expanded
beyond those dates where necessary.
The inspectors also reviewed 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 conclusions reached in
the Wolf Creek corrective action program trending reports.
The inspectors reviewed corrective actions for problem identification and resolution and
human performance cross-cutting themes.
These activities constitute completion of a one semi-annual trend inspection sample as
defined in Inspection Procedure 71152-05.
b. Findings
No findings were identified.
.4 Selected Issue Follow-up Inspection
a. Inspection Scope
During a review of Wolf Creek corrective action program items, the inspectors noted a
condition report documenting over drilling of stud holes on a feedwater regulating valve
body. The inspectors reviewed vendor manuals and station procedures for drilling and
installing Heilicoil inserts. The inspectors reviewed vendor calculations for the strength
of the joint. The inspectors interviewed engineers regarding the procedure and
determined that work performed was consistent with vendor instructions.
- 44 - Enclosure
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.
4OA3 Event Follow-up (71153)
.1 March 19, 2011, Safety Injection
a. Inspection Scope
On March 19, 2011, with the reactor in hot standby, the inspectors responded to the
control room when Wolf Creek received a safety injection signal for a rapid steamline
pressure decrease. The inspectors reviewed control room logs and plant computer data.
The inspectors interviewed control room operators about the conditions leading up to the
event as well as the plant response. The inspectors reviewed plant operating practices
regarding methods of feedwater heating, main steam procedures, and emergency
operating procedures. From interviews with several members of the operating crew and
plant data before and after the event, the inspectors independently reviewed the
sequence of events:
- The crew assumed the watch in Mode 1 and reduced reactor power to Mode 3
for Refueling Outage 18.
- On March 18, 2011, SYS AE-200, Feedwater Preheating During Plant Startup
and Shutdown, Revision 29, was entered for the plant shutdown.
- At midnight, the turbine was tripped in accordance with procedure.
- At 12:07 a.m., feedwater temperature and steam flows begin oscillating. Over
the next hour, feedwater temperature and steam flows oscillated. It was later
determined that the oscillations were due to manual control of FB-PIC 300
combined with solenoid valve air leakage. This action was not peer-checked and
control room supervision was not made aware. PIC-300 controls valve FB-17A
which admits steam to the high pressure feedwater heaters. This is a large
steam demand. The heaters had several temperature swings. This was not
identified until after the safety injection.
- At 12:37 a.m., March 19, the reactor entered Mode 2.
- At 12:54 a.m., March 19, the reactor entered Mode 3.
- At 1:00 a.m., letdown automatically isolated at 17 percent pressurizer level due to
a cooldown in progress.
- 45 - Enclosure
- At 1:01 a.m., reactor coolant temperature could not be maintained, operators
shut the main steam isolation valves to stop the cool down. Reactor coolant
system temperature subsequently recovered to 560°F. Main feedwater pump A
turbine was subsequently tripped from the control room.
temperature decreases.
- With main steam isolation valves shut, the feedwater heaters continued to draw
steam from the main steam header. Steam line temperature decreases.
- At 3:19 a.m., operators re-opened the main steam isolation valve bypass valves.
- To open the main steam isolation valves, operators entered
Procedure SYS AB-120. This procedure is intended for use in Mode 4 with a
maximum steam line pressure of 300 psi. Steam line pressure was
approximately 1000 psi. Precaution 4.5 and step 6.14.2 require that main steam
isolation valve differential pressure be less than 20 psi to open a main steam
isolation valve. Senior reactor operators and management oversight mark these
steps as not applicable.
- At 4:04 a.m., upon opening main steam isolation valve C, a negative steam line
pressure rate on steam line C triggered an automatic safety injection signal.
- Operators entered EMG E-0, Response to Reactor Trip or Safety Injection.
- The pressurizer power-operated relief valve began cycling due to the pressure
increase from the high head centrifugal charging pumps adding inventory to the
- At 4:11 a.m., the safety injection signal was reset and Technical
Specification 3.0.3 was entered for both trains of emergency core cooling system
inoperable because automatic safety injection signal was blocked.
- At 4:12 a.m., high pressure injection was terminated when the boron injection
tank valves were shut.
- At 4:18 a.m., pressurizer power-operated relief valve stops cycling and closes.
- At 4:23 a.m., Wolf Creek transitioned to Procedure EMG ES-03, Safety Injection
Termination.
- At 4:44 a.m., normal letdown flow from the reactor coolant system was re-
established to reduce pressurizer level from a high of 88 percent.
- At 5:20 a.m., Wolf Creek completed Procedure EMG ES-03 and transitioned to
Procedure OFN EM-024 Safety Injection Recovery.
- 46 - Enclosure
- At 5:54 a.m., Wolf Creek notified the headquarters operations officer of the event
by making event notification 46685 per 10 CFR 50.72(b)(2)(iv)(A) which is a
4-hour report for emergency core cooling system discharge into the reactor
coolant system.
- At 6:39 a.m., both reactor trip breakers were closed using SYS SF-120 and
Technical Specification 3.0.3 was exited.
- At 11:21a.m., Wolf Creek updated event notification 46685 with additional
information regarding safety system actuation and loss of an accident mitigation
safety system after the inspectors identified that these 8-hour reporting
requirements may also apply to this event. Condition report 34995 was written
for the potentially missed reports.
- Wolf Creek subsequently left the main steam isolation valves shut and cooled the
plant using the atmospheric relief valves to a temperature where the residual
heat removal system could be placed in service.
b. Findings
Introduction. A Green self-revealing cited violation of Technical Specification 5.4.1.a,
Administrative Procedures, was reviewed involving the failure to correct a previous
violation for an inadequate main steam system procedure. Specifically,
Procedure SYS AB-120 was not corrected to establish appropriate conditions to open a
main steam isolation valve. The inadequate procedure resulted in a safety injection.
Description. The inspectors reviewed a March 5, 2010, event involving excessive steam
generator level swell and feedwater isolation following opening of a main feedwater
isolation valve described in Condition Report 23938 and noncited violation
NCV 05000482/2010004-01. Wolf Creek determined the cause of the March 5, 2010,
P-14 feedwater isolation was an inadequate means of determining the pressure
difference across the main steam isolation valves using control room pressure
indicators. Procedure SYS AB-120, Main Steam and Steam Dump Startup and
Operation, Revision 24, used an acceptance criterion of less than 25 psi differential
pressure to allow opening of a main steam isolation valve. The procedure directed the
operators to determine valve differential pressure using control room indicators before
opening the main steam isolation valves. The control room instruments have ranges
from 0 to 1300 psi or greater with a 25 psi scale and are accurate to within plus or minus
25 to 38 psi. The inspectors concluded the apparent cause evaluation in Condition
Report 23938 appropriately determined that instrument uncertainty equal to or greater
than the procedures acceptance criteria was not reasonable. Subsequently, Wolf Creek
revised Procedure SYS AB-120 to direct operators to determine differential pressure
using locally installed instruments in lieu of the control room pressure indicators, but this
change was only implemented for steam line pressures below 300 psi. Additional
changes were made to several procedures which reduced the allowable steam
generator level band when opening a main steam isolation valve.
Procedure SYS AB-120 revisions did not address steam pressures above 300 psi nor
- 47 - Enclosure
were its precautions and limitations updated to reflect main control board instrumentation
accuracy.
On March 19, 2011, Wolf Creek was in Mode 3 shutting down for a refueling outage.
The steam header pressure was 1000 psi. At 1:01 a.m., operators shut all main steam
isolation valves due to an excessive cooldown. Several hours later, the operators began
to open the valves using Procedure SYS AB-120. A tighter acceptance criterion of
20 psi differential pressure was specified in the procedure before opening a main steam
isolation valve. Wolf Creek operators did not use local instruments as specified by the
procedure. Instead they used control room instruments to determine main steam line
pressures on both side of the main steam isolation valves without considering that the
instrument uncertainty exceeded the range of acceptance criteria. While the control
room pressure and temperature instruments indicated that the differential pressure was
acceptable, actual differential pressure was about 200 psi. When main steam isolation
valve C was opened, a safety injection signal occurred.
The inspectors reviewed corrective actions for Procedure SYS AB-120 and found
several missed opportunities to correct the deficiency. On October 18, 2010, Condition
Report 29168 was written stating Guidance for opening MSIVs not good above 35 psi
steam press, as its problem description. Wolf Creek reviewed Condition Report 30453
which responded to noncited violation NCV 05000482/2010004-01 and appropriately
concluded that the evaluation was flawed for two reasons. First, Condition Report 30453
failed to incorporate the instrument uncertainty issue previously identified in Condition
Report 29168 into the precautions and limitations of Procedure SYS AB-120. Second,
Condition Report 30453 failed to address the full range of anticipated plant conditions
which may require opening a main steam isolation valve, specifically steam pressures
above 300 psi. The inspectors concluded the failure to implement comprehensive
corrective actions to address the March 5, 2010, event directly contributed to the
March 19, 2011, inadvertent safety injection event and constituted a failure to restore
compliance for noncited violation NCV 05000482/2010004-01.
The inspectors reviewed the safety impact of the safety injection transient on Wolf
Creek. Actual safety impacts included a waterhammer on the main steam lines. This
caused a partial failure of main steam isolation valve actuator to bonnet gaskets. The
pressurizer power-operated relief valve 455 cycled seven times. Main feedwater was
lost when the feedwater isolation valves received a close signal from the safety injection.
Emergency core cooling system injection check valve BB8948C experienced body-to-
bonnet gasket leakage. The pressurizer started at 17 percent level and filled to
88 percent level until letdown was reestablished. Inadvertent safety injection has the
potential to challenge the pressurizer safety valves and escalate to a loss of coolant
accident if not terminated.
Analysis. The failure to correct deficiencies in Procedure SYS AB-120 for steam
pressures above 300 psi was a performance deficiency. The inspectors determined that
this finding was more than minor because it impacted the equipment performance
attribute for the Initiating Events Cornerstone and it affected the cornerstone objective to
limit the likelihood of those events that upset plant stability and challenge critical safety
functions during shutdown as well as power operations. Specifically, this issue relates to
- 48 - Enclosure
the configuration control attribute for shutdown equipment alignment. The inspectors
evaluated the significance of this finding using Inspection Manual Chapter 0609.04.
Assuming worst case degradation, the finding resulted in exceeding the technical
specification limit for reactor coolant system leakage due to the pressurizer power-
operated relief valve cycling. Therefore, the inspectors screened the finding to a
Phase 2 review by the senior reactor analyst. The senior reactor analyst used the Wolf
Creek SPAR Model and concluded that the incremental core damage probability
was 3.7E-7, Green. The inspectors found that the cause of the finding has a cross-
cutting aspect in the area of problem identification and resolution associated with the
corrective action program. Specifically, several evaluations failed to include an adequate
extent of condition review that identified that the procedures were inadequate for
opening a main steam isolation valve at system pressures above 300 psi P.1(c).
Enforcement. Technical Specification 5.4.1.a requires that procedures be
established, implemented, and maintained covering the activities described in
Regulatory Guide 1.33, Revision 2, Appendix A. Regulatory Guide 1.33, Revision 2,
Appendix A, Section 3.i, requires procedures for the startup, operation, and
shutdown for the main steam system. Wolf Creek Procedure SYS AB-120, Main
Steam and Steam Dump Startup and Operation, Revision 27, implements these
requirements for the main steam system. Contrary to the above, from March 5,
2010, to March 19, 2011, Wolf Creek Procedure SYS AB-120 had not been
maintained to cover activities for the startup, operation and shutdown of the main
steam system. Specifically, Procedure SYS AB-120, Revision 27, contained
inadequate steps necessary to open a main steam isolation valve without causing a
safety injection signal.This issue and the corrective actions are being tracked by the
licensee in Condition Report 34964. Due to the licensees failure to restore
compliance from previous NCV 05000482/2010004-01 within a reasonable time after
the violation was identified, this violation is being cited as a Notice of Violation
consistent with Section 2.3.2 of the Enforcement Policy: VIO 05000482/2011003-07,
Failure to Correct Procedure for Opening Main Steam Isolation Valves
(EA-11-149).
.2 March 21, 2011, Low Temperature Overpressure System Actuation.
a. Inspection Scope
On March 21, 2011, Wolf Creek was shutdown for a refueling outage. While cleaning
the reactor coolant system, operators failed to maintain reactor coolant system pressure
below 350 psi. When charging was increased for the clean-up, the low temperature
overpressure setpoint was exceeded causing pressurizer power-operated relief
valve 455 to lift three times. The inspectors interviewed reactor operators, reviewed
control room logs, procedures, pressure and temperature limits report, License
Amendment 130, and plant computer data.
- 49 - Enclosure
b. Findings
Introduction. The inspectors reviewed a self-revealing Green noncited violation of
Technical Specification 5.4.1.a, Procedures, for failure to maintain pressure below the
low pressure overpressure protection setpoint.
Description. On March 21, 2011, Wolf Creek was adjusting the chemical and volume
control system to inject hydrogen peroxide into the reactor coolant system to induce a
crud burst to reduce system radioactivity for the refueling outage. Letdown flow was at
approximately 63 gpm. The unit was in Mode 5 with the pressurizer solid and
maintaining reactor coolant system temperature at 160°F and 350 psig pending reactor
coolant system cleanup. The pressurizer is considered solid when it is water filled
because water is not compressible when compared with a gas bubble. Charging and
letdown were in the process of being increased in order to increase the rate of reactor
coolant system cleanup. At 2:52 p.m., power-operated relief valve 455A cycled three
times over the following 4 minutes when reactor coolant system pressure increased to its
lift setpoint of 415 psig. Reactor coolant system pressure control was subsequently
reestablished at 350 psig when letdown flow was increased to approximately 120 gpm.
During interviews, the operators stated that the charging header controller was adjusted
before letdown, and that it was sluggish at the low pressure. The procedure only stated
to maintain pressure and did not provide specific guidance. At the time, operators had a
band of 330-350 psig to maintain, and the operators stated that the normal charging
pump controller was sluggish at its reduced operating pressure. The operators stated
that the charging pump controller was increased three times and on the third time, a
large increase in charging was received.
The inspectors reviewed plant computer data and found that when charging header
pressure was initially increased without increasing letdown flow from the residual heat
removal system, the reactor coolant system pressure rapidly increased. As the 4 minute
event progressed, the normal charging pump controller was adjusted several times while
letdown was progressively increased. Charging header pressure and flow drove the
increases in reactor coolant system pressure. The three lifts of the power operated relief
valve were due to the changes in charging header pressure with a solid pressurizer.
The inspectors reviewed Procedures GEN 00-006 and SYS BG-120 and found that they
did not contain any precautions or limitations regarding the reactor coolant system
pressure response to a sluggish charging controller with a water-solid plant. There were
no instructions that letdown should have been increased first and to adjust charging
second, to match. Procedure GEN 00-006, step 6.44.8.3 only stated to maintain a
pressure band of 325-350 psig when adjusting charging flow. Although the procedures
had several steps to maximize letdown and charging for reactor coolant system clean-
up, there were no specific steps on how to perform this, and there were no continuous
action steps or precautionary steps to prevent over-pressurizing the reactor coolant
system.
The inspectors reviewed the Just in Time Training for the refueling outage and
identified that it contained guidance on raising letdown to 120 gpm and subsequently
taking the plant solid. It did not contain guidance or lessons on manipulating letdown
- 50 - Enclosure
with the plant solid. With the reactor coolant system pressure at the upper end of the
band specified in Procedure GEN 00-006, letdown would be appropriate to adjust first to
prevent the lifting of relief valves. If the reactor coolant system was solid at the lower
end of the pressure band specified in Procedure GEN 00-006, adjusting charging first
would be appropriate to avoid a decrease in reactor coolant system pressure that could
meet the reactor coolant pump trip criteria.
Analysis. Failure to maintain pressure below the power operated relief valve setpoint
was a performance deficiency. The performance deficiency was more than minor
because it impacted the Initiating Events Cornerstone objective of configuration control
to limit the likelihood of those events that upset plant stability and challenge critical
safety functions during shutdown as well as power operations. The significance of the
finding was determined using Inspection Manual Chapter 0609, Significance
Determination Process, Appendix G, Checklist 2, and determined to be of very low
safety significance (Green), because it did not cause the loss of mitigating capability of
core heat removal, inventory control, power availability, containment control, or reactivity
control. Additionally, the finding also did not cause any low temperature overpressure
technical specifications to be exceeded. The inspectors found that the cause of the
finding had a cross-cutting aspect in the area of human performance. Specifically,
operators had to rely on skill of the craft when procedures should have supplied more
instruction for manipulating charging and letdown with the pressurizer water solid H.2.c].
Enforcement. Wolf Creek Technical Specification 5.4.1.a, Procedures, requires, in
part, that written procedures shall be established, implemented and maintained for the
activities recommended in Regulatory Guide 1.33, Revision 2, Appendix A,
February 1978. Regulatory Guide 1.33, Revision 2, Appendix A, February 1978,
General Plant Operating Procedures, Section 2.j, requires procedures for hot standby
to cold shutdown. Procedure GEN 00-006, Hot Standby to Cold Shutdown,
Revision 76, implements this procedure. Procedure GEN 00-006, Step 6.44.8.3 required
the licensee to maintain a pressure band below 350 psig when manipulating charging
flow. Contrary to the above, on March 21, 2011, Wolf Creek did not implement
Procedure GEN 00-006, step 6.44.8.3, to maintain a pressure band below 350 psig
when manipulating charging flow. Because the finding is of very low safety significance
and has been entered into the licensees corrective action program as Condition
Report 35244, this violation is being treated as a noncited violation consistent with
Section 2.3.2 of the NRC Enforcement Policy: NCV 05000482/2011003-08, Failure to
Maintain Reactor Coolant System Pressure Below Relief Valve Setpoint.
.3 April 5, 2011, Vital Switchgear Room Fire
a. Inspection Scope
The inspectors responded to a fire in the switchgear rooms and to the control room. The
inspectors interviewed fire brigade leaders and the control room shift manager regarding
the fire alarms and the fire brigade response and examined the damage inside of
nonvital inverter PN009. The inspectors observed postfire actions to ventilate the area
to remove the smoke and Halon.
- 51 - Enclosure
b. Findings
Introduction. The inspectors reviewed a self-revealing Green noncited violation of
License Condition 2.C.5 for failure to implement adequate fire impairments which
affected both trains of vital ac and dc switchgear.
Description. On March 26, 2011, Wolf Creek implemented a breach authorization
requiring a continuous fire watch because the doors between vital ac and dc switchgear
rooms were propped open. These doors are 3-hour fire barriers. This was done to allow
the train B air conditioning unit and ventilation to provide cooling to the train A
switchgear in accordance with Procedure SYS GK-200, Inoperable Class IE A/C Unit.
With the train A air conditioning unit out of service, two sets of double doors were
propped open between vital ac switchgear trains A and B on the 2000 elevation of the
control building. On April 5, 2010, Wolf Creek completed preventive maintenance on
nonvital inverter PN009 which is located in the 2000 elevation train A vital switchgear
room. Wolf Creek was preparing to test nonvital inverter PN009 and reenergized it for
about 20 minutes. Two electricians were at the inverter cabinet in the train A vital
switchgear room when smoke began emanating from the top of the cabinet. The
electricians shut off the dc input and opened the ac output breakers on the lower door of
the cabinet. The Halon actuation alarm sounded indicating that Halon would discharge
into the room in 30 seconds. One electrician told the fire watch that it was necessary to
evacuate. The two electricians and the fire watch were egressing through the north
missile door when the Halon system discharged. The breached doors between ac
switchgear rooms were not shut. The fire brigade responded and removed an extension
cord and shut the doors between the vital ac switchgear rooms. The fire brigade found
only smoke and Halon in the rooms and no fire at PN009. Subsequent examination by
Wolf Creek and the vendor found that vendor errors in labeling the terminals caused an
excessive current in an adjacent transformer which caused the fire. Both transformers
were replaced. The vendor stated that no other damage occurred. Condition
Report 36719 was written on the inadequate fire watch response.
The inspectors interviewed the April 5 fire watch and found that he thought Halon was
going to discharge into both the trains A and B vital switchgear rooms. Thus, he would
have to egress through the north missile door and not to the train B switchgear room.
He understood his duty to shut the doors upon alarm, but indicated that he would not be
able to remove the extension cord, shut the doors, and exit within 30 seconds. The fire
watch stated that removing the extension cord and shutting the doors would likely take
3 to 4 minutes. The inspectors found that the Halon system was designed to discharge
into the switchgear room with the alarming smoke detectors. The fire watch also stated
that he left the room without shutting the doors because the electricians instructed him to
leave the room prior to the Halon actuation.
The inspectors found that the only written instructions for fire watches was the statement
on the fire impairment which said Per AP 10-104, section 6.1.9 (SYS GK-200), in case
of fire or Halon discharge, close doors 33011 & 33023 and exit area and notify control
room. Wolf Creek relies on training and reading of the fire impairment to understand
the compensatory action. The inspectors reviewed the design of the 1301 Halon system
and found that the system was sized to extinguish a fire in one switchgear room only.
- 52 - Enclosure
The inspectors found that with doors 33011 and 33023 (each a set of double doors)
open between vital switchgear rooms, the Halon system would not have been successful
at extinguishing a fire.
The inspectors reviewed written statements from the fire brigade, the fire watch, and the
electricians. The inspectors reviewed Procedure AP 10-107, Fire Protection Impairment
Control, and Procedure APF 10-104-01, Breach Authorization Permit, and found that
the requirements of the breach permit were not met because the fire watch failed to
close doors 33011 and 33023 during an actual fire. Procedure AP 10-104, step 5.62,
states, in part, that the boundary watch must be able to clear any cord or tool crossing a
breached barrier and to notify the control room if any condition in which a breached
barrier cannot be closed within the time requirements. The inspectors reviewed form
APF 10-104-01, breach authorization, for the 2000 and 2016 elevation switchgear
rooms and found no quantitative timing requirements for closing the doors. The
inspectors concluded that a 30-second acceptance criterion was critical because open
doors would prevent the Halon system from reaching the necessary concentration to
extinguish a fire. The inspectors found that the breach permit was not met because the
fire watch did not close the doors and that the breach permit was inadequate because it
did not contain timed acceptance criteria necessary to ensure the success of the Halon
system.
On April 12, 2011, the inspectors interviewed a different vital ac switchgear room fire
watch and found that the watch was also not clear on their duty to shut the doors
regardless of what other workers tell them to do. The fire watch did have correct
knowledge of the Halon system, the 30-second delay between the alarm and discharge,
and what room to egress to depending on the fire location. The inspectors shared this
with the outage control center.
On April 14, 2011, Wolf Creek inhibited the Halon systems for the rod-drive motor
generator set room, all vital dc switchgear rooms, and the vital ac switchgear rooms.
Wolf Creek judged it more important to ensure that the fire watches shut the fire doors
rather than have an ineffective automatic Halon system actuation. On April 13, 2011,
Wolf Creek initiated new training for all fire watches to ensure they had copies of the
breach permits. The inspectors interviewed fire watches on April 14, 2011, and found
that the watches had a complete and correct understanding of their duties.
The inspectors found that the inability to close these fire doors was identified in
Refueling Outage 16 in Condition Report 2008-1357. Actions included protective
equipment and training to remove cables crossing the open doors, but the inspectors
concluded that those corrective actions did not ensure proper fire watch actions on
April 5, 2011. Corrective actions included APF 10-104-06 to include Special
Requirements for Boundary Watch. Although a new section of the breach impairments
was created, it was typically not utilized when breaching vital switchgear doors.
As an extent of condition review, the inspectors reviewed previous fire impairments for
Procedure SYS GK-200 for open fire doors on the 2000 and 2016 control building
elevations. The inspectors found that Procedure SYS GK-200 had been implemented
- 53 - Enclosure
23 times over the prior year of operation representing approximately 36 days of
impairments for both trains of ac and dc switchgear and batteries.
Analysis. The failure to implement adequate fire watches that ensured the success of
the Halon system was considered a performance deficiency. The performance
deficiency was considered more than minor because it impacted the Initiating Events
Cornerstone and its objective to limit the likelihood of those events that upset plant
stability and challenge critical safety functions during shutdown as well as power
operations. Specifically, the fire area of the protection against external factors attribute
was impacted. The inspectors used Inspection Manual Chapter 0609.04 to screen the
finding to Inspection Manual Chapter 0609, Appendix F, because the fire protection
defense-in-depth strategies involving automatic suppression, fire barriers, administrative
controls were degraded. Because the subject finding was not clearly covered by the
approach used in Appendix F, the senior reactor analyst performed a Phase 3 analysis.
The doors were open for 36 days, so a 36-day exposure period (EXP) was used. The
analyst used generic values for the Fire Ignition Frequency (FI), Severity Factor (PSF)
and the probability of manual suppression before damage (PMS). The Fire Mitigation
Frequency (FM) was calculated as follows:
FM = FI * PSF * PMS * EXP
= 2.0 x 10-2/year * 0.1 * 0.1 * 36 days ÷ 365 days/year
= 1.97 x 10-5
The analyst assumed that if the fire grew to a point that it could spread to the opposite
train, it would actuate the opposite trains Halon system and cause an isolation of all
ventilation. However, there was no credible source of flammable materials that would
cause the growth of the fire into the opposite trains switchgear. Therefore, the analyst
quantified the conditional core damage probability (CCDP) for the failure of
Switchgear NB01 using the Standardized Plant Analysis Risk Model for Wolf Creek
Station, Revision 8.15. The resulting CCDP was 8.3 x 10-4. The final change in core
damage frequency (CDF) was calculated as follows:
= 1.97 x 10-5 * 8.3 x 10-4
= 1.6 x 10-8
Therefore, this finding was determined to be of very low safety significance (Green).
The inspectors found that the cause of the finding had a cross-cutting aspect in the area
of problem identification and resolution. Specifically, corrective actions from 2008
ineffective fire watches did not prevent recurrence of the April 5, 2011, inadequate fire
watch P.1.d].
Enforcement. License condition 2.C.(5) states, in part, that the licensee shall maintain in
effect all provisions of the approved fire protection program as described in the
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Standardized Nuclear Unit Power Plant System USAR for the facility through
Revision 17, the Wolf Creek site addendum through Revision 15, and as approved in the
safety evaluation report through Supplement 5, Amendments 191 and 193. AP 10-100,
fire protection program, states, in part, that AP 10-104, Breach Authorization, is part of
the fire protection program. Procedure AP 10-104, step 5.62, states, in part, that the
boundary watch must be able to clear any cord or tool crossing a breached barrier and
to notify the control room if any condition in which a breached barrier cannot be closed
within the time requirements. Procedure AP 10-104, steps 6.1.8 and 6.1.9, require, in
part, that a continuous fire watch shall be established for the vital switchgear rooms
because open doors will reduce Halon concentration and expose redundant trains to the
same fire. Contrary to the above, prior to April 14, 2011, the licensee failed to implement
and maintain in effect all provisions of the approved fire protection program. Specifically,
the licensee used an ineffective fire barrier breach permit system that did not ensure that
the Halon systems would effectively extinguish fires because the fire watches could not
clear any cord or tool crossing a breached barrier and did not notify the control room of a
condition in which a breached barrier could not be closed within the time requirements.
The licensee entered this issue into their corrective action program as Condition
Report 36719. Because this violation was of very low safety significance and it was
entered into the corrective action program, this violation is being treated as a noncited
violation, consistent with the NRC Enforcement Policy, Section 2.3.2:
NCV 05000482/2011003-09, Inadequate Fire Watch Defeats Halon Fire Suppression in
Vital Switchgear Rooms During Fire.
.4 (Closed) Licensee Event Report (LER) 2006-003-00, Indications Discovered on
Pressurizer during Preplanned Inservice Inspections
On October 11, 2006, during Refueling Outage 15, engineering personnel performing
preplanned inservice examination of the pressurizer nozzle to safe end dissimilar metal
welds identified five circumferential flaw indications. Three indications were located in
the surge nozzle dissimilar metal weld, one indication was in the safety nozzle C
dissimilar metal weld, and one indication was in the relief nozzle dissimilar metal weld.
The locations were all part of the reactor coolant system pressure boundary. There was
no evidence of reactor coolant system pressure boundary leakage. The most probable
mechanism responsible for the indications was primary water stress corrosion cracking.
Wolf Creek Generating Station was in Mode 5, cold shutdown. Weld overlay repairs of
the flaw indications were performed prior to the unit's return to power operations. The
inspectors reviewed LER 05000482/2006-003-00 to verify that the cause was identified
and that corrective actions were appropriate. This LER is closed.
.5 (Closed) Notice of Violation VIO 05000482/2010006-05, Failure to Correct NRC
Identified NCV Apparent Cause Evaluation Vice Root Cause Evaluation for Essential
The violation involved the failure to perform an adequate cause evaluation and to take
corrective actions to preclude repetition for a significant condition adverse to quality.
Although determined to be of very low safety significance (Green), this violation was
cited in Notice of Violation 05000482/2010006-05 because not all of the criteria specified
in Section 2.3.2 of the NRC Enforcement Policy were satisfied (EA-10-160). Specifically,
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the Wolf Creek Generating Station failed to restore compliance within a reasonable time
for a previously NRC identified noncited violation as documented in NRC Inspection
Report 05000482/2009007-03. The inspectors reviewed the corrective actions
completed by the licensee and verified that the cause was identified and that corrective
actions were appropriate. This violation is closed.
4OA5 Other Activities
.1 (Closed) NRC TI 2515/177, Managing Gas Accumulation in Emergency Core Cooling,
Decay Heat Removal, and Containment Spray Systems (NRC Generic Letter 2008-01)
a. Inspection Scope
As documented in Sections 1R04.1 and 1R22 of this report, the inspectors confirmed the
acceptability of the described actions for the high pressure safety injection system and
the containment spray system. This inspection effort counts towards the completion of
TI 2515/177 which is closed in this inspection report.
The inspectors evaluated whether the licensee maintained documents, installed system
hardware, and implemented actions with the information provided in their response to
NRC Generic Letter 2008-01, Managing Gas Accumulation in Emergency Core Cooling,
Decay Heat Removal, and Containment Spray Systems. Specifically, the inspectors
verified that the licensee has implemented or was in the process of implementing the
commitments, modifications, and programmatically controlled actions described in the
response to Generic Letter 2008-01. The inspectors conducted their review in
accordance with TI 2515/177 and considered the site-specific supplemental information
provided by the Office of Nuclear Reactor Regulation (NRR) to the inspectors.
b. Inspection Documentation
The inspectors reviewed the licensing basis, design, testing, and corrective actions as
specified in the TI. The specific items reviewed and any resulting observations are
documented below.
Licensing Basis. The inspectors reviewed selected portions of licensing basis
documents to verify that they were consistent with the NRR assessment report and that
the licensee properly processed any required changes. The inspectors reviewed
selected portions of technical specifications, technical specification bases, and the
USAR. The inspectors also verified that applicable documents that described the plant
and plant operation, such as calculations, piping and instrumentation diagrams,
procedures, and corrective action program documents, addressed the areas of concern
and were changed, if needed, following plant changes. The inspectors confirmed that
the licensee performed surveillance tests at the frequency required by the technical
specifications. The inspectors verified that the licensee tracked their commitment to
evaluate and implement any changes that will be contained in the technical specification
task force traveler.
- 56 - Enclosure
Design. The inspectors reviewed selected design documents, performed system
walkdowns, and interviewed plant personnel to verify that the licensee addressed design
and operating characteristics. Specifically:
- The inspectors verified that the licensee had identified the applicable gas
intrusion mechanisms for their plant.
- The inspectors verified that the licensee had established void acceptance criteria
consistent with the void acceptance criteria identified by NRR. If NRR
acceptance criteria were not met, then the inspectors verified that the licensee
has justified the deviations. The inspectors also confirmed that the range of flow
conditions evaluated by the licensee was consistent with the full range of design
basis and expected flow rates for various break sizes and locations.
The inspectors noted that the licensee used the methods developed by
Westinghouse to estimate the suction voids emergency core cooling system
pumps. Westinghouse documented their review and test results
in WCAP-16631-NP, Testing and Evaluation of Gas Transport to the Suction of
ECCS Pumps. Wolf Creek used WCAP-16631-NP to show that GOTHIC can
acceptably predict quantitative void transport behavior. However, the inspectors
noted that test configuration and conditions differed from actual plant
configuration and conditions. These methods relied on industry testing
documented by Westinghouse and used the GOTHIC computer code to better
estimate the impacts resulting from voiding in the emergency core cooling
systems.
The licensee had received analyses for their facility based upon the simplified
equation developed by Westinghouse, which would more accurately estimate the
void sizes allowed on the suction of the emergency core cooling pumps without
affecting operability. In addition, the license had received a revised estimate of
water hammer effects developed by Fauske on the pump discharges for their
emergency core cooling systems. These analyses would replace the use of
GOTHIC. These analyses allow for a more realistic estimate of void sizes on
both the suction and discharge of the emergency core cooling system pumps.
The licensee had not accepted these analyses at the time of this inspection.
The inspectors discussed with NRR that the licensee had used these methods.
The ultimate acceptability of these methods required further evaluation by NRR
to: (1) better understand the acceptability of the application of the revised test
results contained in WCAP-16631-NP to void assessment analysis; (2) better
understand and evaluate the use of the simplified equation; and (3) assess
potential generic implications. The licensee documented these outstanding
issues in Condition Report 39943.
- The inspectors selectively reviewed applicable documents, including calculations,
and engineering evaluations with respect to gas accumulation in the emergency
core cooling systems. Specifically, the inspectors verified that these documents
addressed venting requirements, aspects where pipes were normally void such
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as some spray piping inside containment, void control during maintenance
activities, and the effect of debris on strainers in containment emergency sumps
causing accumulation of gas under the upper elevation of strainers and the
impact on the required net positive suction head.
- The inspectors conducted a walk down of selected regions of the emergency
core cooling systems in sufficient detail to assess the licensees walk downs.
The inspectors completed a full containment spray system alignment as
documented in Section 1R04. The inspectors also verified that the information
obtained during the licensees walkdown was consistent with the items identified
during the inspectors independent walk down.
- The inspectors verified that piping and instrumentation diagrams and isometric
drawings that describe the residual heat removal and safety injection system
configurations. The review of the selected portions of isometric drawings
considered the following:
1. High point vents were identified.
2. High points without vents were recognizable.
3. Other areas where gas could accumulate and potentially impact
operability, such as at orifices in horizontal pipes, isolated branch lines,
heat exchangers, improperly sloped piping, and under closed valves,
were described in the drawings or in referenced documentation.
4. Horizontal pipe centerline elevation deviations and pipe slopes in
nominally horizontal lines that exceed specified criteria were identified.
5. All pipes and fittings were clearly shown.
6. The drawings were up to date with respect to recent hardware changes,
and that any discrepancies between as-built configurations and the
drawings were documented and entered into the corrective action
program for resolution.
- The inspectors verified that the licensee had completed their walkdowns and
selectively verified that the licensee identified discrepant conditions in their
corrective action program and appropriately modified affected procedures and
training documents. The inspectors determined that the licensee appropriately
considered the differing gas intrusion mechanisms with one exception. The
inspectors noted that the licensee failed to analyze whether vortexing would
occur in their containment spray additive tank. The details of this issue are
described in Section 4OA5.1.e of this report.
Testing. The inspectors reviewed selected surveillance, postmodification test, and
postmaintenance test procedures and results implemented during power and shutdown
operations to verify that the licensee had approved and used procedures that addressed
- 58 - Enclosure
gas accumulation and/or intrusion into the subject systems. This review included the
verification of procedures used for conducting surveillances and determination of void
volumes to ensure that the licensee satisfied the established void criteria with
reasonable assurance until the next scheduled void surveillance. Also, the inspectors
reviewed procedures used for filling and venting following conditions that may have
introduced voids into the subject systems to verify that the procedures addressed testing
for such voids and provided processes for their reduction or elimination. The inspectors
observed the performance of the emergency core cooling system void surveillance as
documented in Section 1R22.
Corrective Actions: The inspectors reviewed selected actions from the 2011
assessment review and sampled other corrective action program documents to assess
how effectively the licensee addressed the issues in their corrective action program
associated with Generic Letter 2008-01. In addition, the inspectors verified that the
licensee implemented appropriate corrective actions for condition reports identified in the
9-month and supplemental responses. The inspectors determined that the licensee had
initiated a large number of corrective actions in response to previous events at their
facility. The inspectors determined that the licensee had effectively implemented the
actions required by Generic Letter 2008-01.
Based on this review, the inspectors concluded that reasonable assurance exists the
licensee will continue to implement the requirements of Generic Letter 2008-01 and will
complete all outstanding items. This TI is closed.
1. Findings
Introduction. The inspectors identified a Green noncited violation of 10 CFR Part 50,
Appendix B, Criterion III, Design Control for the failure to translate the design basis into
instructions, procedures, and drawings. The inspectors found that the licensee failed to
assess whether vortexing occurred in the containment spray additive tank during a
design-basis accident.
Description. The inspectors evaluated licensee activities related to evaluation of gas
intrusion into their emergency core cooling systems. The inspectors questioned whether
air entrainment in the containment spray system, as a result of vortexing in the
containment spray additive tank, affected the ability of the containment spray system to
remain full of water and meet the accident flow requirements. The licensee did not have
a calculation to determine whether vortexing would occur in their containment spray
additive tank at the required design flow rates. The licensee initiated Condition
Report 38715 to document this deficiency; initiated actions to calculate the effects of
vortexing in the containment spray additive tank during design basis flows; and
established a Mode 3 restraint related to completing the calculation to ensure that
containment spray would be operable as required.
The system used an eductor driven by discharge flow from both of the containment
spray pumps to draw sodium hydroxide from the single chemical additive tank during a
design-basis accident. Vacuum breakers allowed air into the tank as the liquid drained.
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Calculation EN M-024, Critical Submergence in Containment Spray Additive
Tank (TEN01) to Avoid Vortex, Revision 0, concluded that vortexing would not occur.
Analysis. Failure to implement design control measures to analyze whether containment
spray piping remained full of water was a performance deficiency. This finding was
more than minor because it affected the design control attribute of the Mitigating
Systems Cornerstone objective to ensure the availability, reliability, and capability of the
containment spray system to respond to initiating events and prevent undesirable
consequences. Specifically, the inspectors had reasonable doubt on the capability of
the containment spray system to properly inject because of vortexing in the containment
spray additive tank. The inspectors performed the significance determination using
Inspection Manual Chapter 0609.04. The finding was determined to be of very low
safety significance (Green) because it was a design or qualification deficiency confirmed
not to result in loss of operability or functionality. Although the failure to have this
calculation had existed since original construction, the inspectors determined this finding
reflected current performance since the licensee was required to evaluate likelihood of
tanks allowing gas intrusion into the emergency core cooling systems in response to
Generic Letter 2008-01, Managing Gas Accumulation in Emergency Core Cooling,
Decay Heat Removal, and Containment Spray Systems. Consequently, this finding had
problem identification and resolution cross-cutting aspects associated with the corrective
action program in that the licensee did not evaluate thoroughly the potential for gas
intrusion from all possible tanks P.1(c).
Enforcement. Title 10 CFR Part 50, Appendix B, Criterion III, requires, in part, that
design control measures shall provide for verifying or checking the adequacy of design,
such as by the performance of design reviews, by the use of alternate or simplified
calculational methods, or by the performance of a suitable testing program, was
identified. Specifically, the design capability of the containment spray system requires
that the system be full of water in order to achieve and maintain the design rate of flow.
Contrary to the above, as of May 6, 2011, the licensee had not verified the adequacy of
the design capability of the containment spray system to remain full of water through
design review, calculation, or testing. Specifically, the licensee had not analyzed
whether vortexing in the containment spray additive tank would affect system flow. The
analysis demonstrated that no air should be entrained as a result of vortexing. The
licensee documented this issue in Condition Report 38715. Because this finding was of
very low safety significance and has been entered into the corrective action program, it
is being treated as a noncited violation consistent with Section 2.3.2 of the NRC
Enforcement Policy: NCV 05000482/2011003-10, Failure to Analyze for Vortexing in
Containment Spray Additive Tank.
.2 (Closed) NRC TI 2515/183, Followup to the Fukushima Daiichi Nuclear Station Fuel
Damage Event
a. Inspection Scope
The inspectors assessed the activities and actions taken by the licensee to assess its
readiness to respond to an event similar to the Fukushima Daiichi nuclear plant fuel
damage event. This included (1) an assessment of the licensees capability to mitigate
- 60 - Enclosure
conditions that may result from beyond design basis events, with a particular emphasis
on strategies related to the spent fuel pool, as required by NRC Security Order
Section B.5.b issued February 25, 2002, as committed to in severe accident
management guidelines, and as required by 10 CFR 50.54(hh); (2) an assessment of
the licensees capability to mitigate station blackout conditions, as required by
10 CFR 50.63 and station design bases; (3) an assessment of the licensees capability
to mitigate internal and external flooding events, as required by station design bases;
and (4) an assessment of the thoroughness of the walkdowns and inspections of
important equipment needed to mitigate fire and flood events, which were performed by
the licensee to identify any potential loss of function of this equipment during seismic
events possible for the site.
b. Findings and Observations
NRC Inspection Report 05000482/2011008 (ML11133A354) documented detailed results
of this inspection activity. Following issuance of the report, the inspectors conducted
additional follow-up on the following seven selected issues.
1. Extensive damage mitigation guideline procedures specify that if the control room
staff and field operators are compromised, then the shift security commander
becomes the incident coordinator until an operator can be found. The inspectors
identified that shift security commanders are not trained on reactor technology
and mitigating systems, therefore it is not reasonable to assume they would have
a sufficient knowledge base or decision making ability to direct technical
response to an extensive damage situation. The licensee entered the issue into
their corrective action program and is in the process of conducting additional
procedural and technical training for security commanders.
The inspectors reviewed licensee extensive damage mitigation guidelines in
greater detail and compared them to the requirements of 10 CFR 50.54(hh)(2) as
well as to the expectations outlined in the NRC Staff Guidance for Use in
Achieving Satisfactory Compliance with February 25, 2002, Order Section B.5.b
dated February 25, 2005, and determined that Wolf Creeks procedures meet
agency expectations in that they direct security commanders to seek out persons
with the best technical expertise available. This issue of concern is closed with
no finding.
2. The licensee identified that extensive damage mitigation guidelines procedures
to refill the refueling water storage tank are not viable because the connection
point is not readily accessible. The licensee entered this issue into their
corrective action program and is evaluating potential design changes to resolve
this concern.
The inspectors reviewed the applicable extensive damage mitigation attachments
which direct refilling of the refueling water storage tank in greater detail and
compared them to the requirements of 10 CFR 50.54(hh)(2) as well as to the
expectations outlined in the NRC Staff Guidance for Use in Achieving
Satisfactory Compliance with February 25, 2002, Order Section B.5.b, dated
- 61 - Enclosure
February 25, 2005, and determined that these procedures do not meet regulatory
requirements for compliance with Order 10 CFR 50.54(hh)(2) because the
expectation element could not be effectively implemented using existing or
readily available resources and because personnel safety concerns associated
with the expectation element had not been addressed. There is no guidance as
to how the connection is to be accessed, nor is the required equipment needed
access and work safely at heights pre-staged in advance. This issue of concern
is documented as a licensee identified violation in Section 4OA7.1 of this report.
3. The licensee identified that extensive damage mitigation guideline procedures
require additional precautionary guidance to prevent excessive reactor coolant
system depressurization which could compromise natural circulation core
cooling. The licensee entered this issue into their corrective action program and
is evaluating procedural enhancements to remedy this concern.
The inspectors reviewed the applicable licensee extensive damage mitigation
attachments which direct actions which can cool and depressurize the reactor
coolant system and determined that this issue of concern was an enhancement
only and not a violation of regulatory requirements. Since operators reviewing
these procedures identified the same concerns and because the licensee has
entered this issue in their corrective action program this issue of concern is
closed with no finding.
4. The licensee identified that alternate power sources specified by extensive
damage mitigation guidelines procedures are not properly staged in advance.
Additional technical guidance on the configuration and use of these sources
needs to be added to the extensive damage mitigation guidelines procedures.
The licensee entered this issue into their corrective action program and is
evaluating alternative equipment and procedural enhancements to resolve this
concern.
The inspectors reviewed the applicable licensee extensive damage mitigation
attachments which direct the use of alternate dc sources in greater detail and
compared them to the requirements of 10 CFR 50.54(hh)(2) as well as to the
expectations outlined in the NRC Staff Guidance for Use in Achieving
Satisfactory Compliance with February 25, 2002, Order Section B.5.b dated
February 25, 2005, and determined that these procedures do not meet regulatory
requirements for compliance with 10 CFR 50.54(hh)(2) because the expectation
element could not be effectively implemented using existing or readily available
resources. Specifically, the components are not pre-staged in advance. This
issue of concern is documented as a licensee identified violation in
Section 4OA7.1 of this report.
5. During walkdowns with the inspector, nuclear station operators failed to promptly
locate certain station blackout emergency operating procedure components in
the plant. The inspectors determined that this was due to inadequate training,
lack of specific procedural guidance, and over-reliance on a computer database
of equipment locations. The computer database would be unavailable during an
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actual station blackout. The licensee agreed with this characterization and
entered this issue into their corrective action program.
6. The inspectors determined that this issue of concern was a performance
deficiency and a violation of Technical Specification 5.4.1.a which requires, in
part, that specified station procedures be established, implemented, and
maintained. The inspectors determined that all of the components operators
failed to identify for local actions were backed up by components which would fail
safe in a loss of ac power event and therefore did not have the potential to further
complicate that event. The inspectors evaluated the issue using Inspection
Manual Chapter 0612, Appendix B, Issue Screening, and determined the failure
to comply with Technical Specification 5.4.1.a constituted a violation of minor
significance that is not subject to enforcement action in accordance with the
NRCs Enforcement Policy. The inspectors also found that Wolf Creek had
completed appropriate corrective actions in this area. This issue of concern is
closed as an NRC identified minor violation.
7. The licensee identified that some fire protection equipment is not stored in
seismic or tornado qualified locations. The licensee identified that the water
supply pumps and piping used for fire protection and extensive damage
mitigation guideline actions is not seismic or tornado qualified. The licensee also
identified that equipment used to access underground diesel storage tanks is not
seismic or tornado qualified; also the tanker truck used to refill the diesel-driven
fire pump and fire truck is not parked in a seismic or tornado qualified building.
The licensee entered these issues into their corrective action program.
8. The inspectors reviewed the requirements of 10 CFR 50.54(hh)(2) as well as to
the expectations outlined in the NRC Staff Guidance for Use in Achieving
Satisfactory Compliance with February 25, 2002, Order Section B.5.b, dated
February 25, 2005, and determined that those regulatory requirements apply only
to fire and explosion events, not to earthquakes and tornadoes. Because Wolf
Creek identified this issue and entered into their corrective action program and
because this issue of concern has no associated violation of regulatory
requirements, it does not meet the criteria of a finding under the Inspection
Manual Chapter 0612. This issue of concern is closed with no finding.
9. The condensate storage tank used in station blackout response and extensive
damage mitigation guideline procedures is not seismic or tornado qualified. The
licensee entered the issue into their corrective action program. The inspectors
found that the safety-related source, from the essential service water system,
would not be impacted. The inspectors reviewed applicable sections of Wolf
Creeks USAR and determined that this issue is within the boundaries of Wolf
Creeks NRC-approved design bases. This issue of concern is closed with no
finding.
- 63 - Enclosure
.3 (Closed) NRC TI 2515/184, Availability and Readiness Inspection of Severe Accident
Management Guidelines (SAMGs)
The inspectors reviewed the licensees severe accident management guidelines
(SAMGs), implemented as a voluntary industry initiative in the 1990s, to determine
(1) whether the SAMGs were available and updated, (2) whether the licensee had
procedures and processes in place to control and update its SAMGs, (3) the nature and
extent of the licensees training of personnel on the use of SAMGs, and (4) licensee
personnels familiarity with SAMG implementation.
The results of this review were provided to the NRC task force chartered by the
Executive Director for Operations to conduct a near-term evaluation of the need for
agency actions following the Fukushima Daiichi fuel damage event in Japan. Plant-
specific results for Wolf Creek were provided as Enclosure 14 to a memorandum to the
Chief, Reactor Inspection Branch, Division of Inspection and Regional Support, dated
May 27, 2011 (ML111470264).
.4 (Closed) TI 2515/172, Reactor Coolant System Dissimilar Metal Butt Welds
a. Inspection Scope
Portions of TI 2515/172 were previously performed at Wolf Creek Nuclear Generating
Station, during Refueling Outages 15, 16, and 17. The results of those inspections are
documented in NRC Inspection Reports 05000482/2006005, 05000482/2008003,
05000482/2009005 and 05000482/2011003, respectively. Specific documents reviewed
during this inspection are listed in the attachment. This unit has the following dissimilar
metal butt welds:
COMPONENT ID DESCRIPTION MRP-139 BASELINE EXAM
CATEGORY
RV-301-121-A Loop 1 Outlet D April 2005 RF14
Nozzle to Safe-
End Weld
RV-301-121-B Loop 2 Outlet D April 2005 RF14
Nozzle to Safe-
End Weld
RV-301-121-C Loop 3 Outlet D April 2005 RF14
Nozzle to Safe-
End Weld
RV-301-121-D Loop 4 Outlet D April 2005 RF14
Nozzle to Safe-
End Weld
- 64 - Enclosure
COMPONENT ID DESCRIPTION MRP-139 BASELINE EXAM
CATEGORY
RV-302-121-A Loop 1 Inlet E April 2005 RF14
Nozzle to Safe-
End Weld
RV-302-121-B Loop 2 Inlet E April 2005 RF14
Nozzle to Safe-
End Weld
RV-302-121-C Loop 3 Inlet E April 2005 RF14
Nozzle to Safe-
End Weld
RV-302-121-D Loop 4 Inlet E April 2005 RF14
Nozzle to Safe-
End Weld
TBB03-1-W Pressurizer F October 2006 RF15
/MW7090-WOL-DM Surge Nozzle to
Safe-End Weld
TBB03-2-W Pressurizer B October 2006 RF15
/MW7089-WOL-DM Spray Nozzle to
Safe-End Weld
TBB03-3-A-W Pressurizer B October 2006 RF15
/MW7086-WOL-DM Safety Nozzle A
to Safe-End
TBB03-3-B-W Pressurizer B October 2006 RF15
/MW7087-WOL-DM Safety Nozzle B
to Safe-End
TBB03-3-C-W Pressurizer F October 2006 RF15
/MW7088-WOL-DM Safety Nozzle C
to Safe-End
TBB03-4-W Pressurizer F October 2006 RF15
/MW7085-WOL-DM Relief Nozzle to
- 65 - Enclosure
COMPONENT ID DESCRIPTION MRP-139 BASELINE EXAM
CATEGORY
Safe-End Weld
1. Licensees Implementation of the Materials Reliability Program (MRP-139) Baseline
Inspections (03.01)
The inspectors reviewed records of structural weld overlays and nondestructive
examination activities associated with the licensees pressurizer and hot leg
structural weld overlay mitigation effort. The baseline inspections of the pressurizer
dissimilar metal butt welds were completed as noted in the table above. The
pressurizer dissimilar metal butt welds had full structural weld overlay applied in
Refueling Outage 15. The first Component ID in the preceding table was the
designation prior to the overlay; the latter Component ID is the current weld
designation (after overlay).
The licensee requested the deviations from the MRP-139 baseline inspection
requirements. These locations are now examined in accordance with the approved
alternative of relief request I3R-05. The licensee did not take any other deviations
from the baseline inspection requirements of MRP-139, and all other applicable
dissimilar metal butt welds were scheduled in accordance with MRP-139 guidelines.
2. Volumetric Examinations (03.02)
The results of these inspections are documented in NRC Inspection
Reports 05000482/2006005, 05000482/2008003, and 05000482/2009005.
3. Weld Overlays (03.03)
Only the pressurizer nozzles have been mitigated. The mitigation type was full
structural weld overlay applied in Refueling Outage 15. A pre-service exam in
accordance with relief request I3R-05 was performed. An inservice exam on the
MRP-139, category F welds was performed in Refueling Outage 16 in accordance
with I3R-05. This examination also falls within the guidelines of MRP-139 for
category F welds.
4. Mechanical Stress Improvement (03.04)
The licensee did not employ a mechanical stress improvement process.
5. Inservice Inspection Program (03.05)
The licensee has prepared an MRP-139 inservice inspection program. All the welds
in the MRP-139 inservice inspection program are appropriately categorized in
accordance with MRP-139. The inservice inspection frequencies are consistent with
the inservice inspection frequencies called for by MRP-139.
- 66 - Enclosure
b. Findings
No findings were identified.
4OA6 Meetings
Exit Meeting Summary
On April 1, 2011, the inspectors presented the inservice inspection results to Mr. S. Hedges, Site
Vice President, and other members of the licensee staff. The licensee acknowledged the issues
presented. The inspectors telephonically re-exited with Mr. Hedges, Site Vice President, and
other members of the licensees staff on June 16, 2011. The inspectors acknowledged review
of proprietary material during the inspection which was returned to the licensee.
On April 5, 2011, the Deputy Director of the Division of Reactor Projects conducted a regulatory
performance meeting in conjunction with the public annual assessment meeting with
Mr. M. Sunseri, President and Chief Executive Officer, and other members of the licensee staff
to review the corrective actions related to the previously White performance indicators for
unplanned scrams per 7000 critical hours, unplanned scrams with complications, and safety
system functional failures.
On May 6, 2011, the inspectors presented the inspection results to Mr. M. Sunseri, President
and Chief Executive Officer, and other members of the licensee staff. The licensee
acknowledged the issues presented. The inspectors confirmed that none of the potential report
input discussed was considered proprietary.
On June 8, 2011, the inspector communicated the results of the in-office inspection of changes
to the licensees emergency plan to Mr. T. East, Superintendent, Emergency Planning, and to
Mr. W. Muilenburg, Licensing Engineer, of the licensees 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 July 13, 2011, the inspectors presented the inspection results to Mr. S. Hedges, Site Vice
President, 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. Although proprietary information was used during
the inspection, it was returned to the licensee or destroyed.
4OA7 Licensee-Identified Violations
The following violations of very low safety significance (Green) were identified by the licensee
and are violations of NRC requirements which meet the criteria of Section 2.3.2 of the NRC
Enforcement Policy for being dispositioned as noncited violations.
.1 Title 10 CFR 50.54(hh)(2)(ii) states: Each licensee shall develop and implement
guidance and strategies intended to maintain or restore core cooling,
containment, and spent fuel pool cooling capabilities under the circumstances
associated with loss of large areas of the plant due to explosions or fire, to
- 67 - Enclosure
include strategies in the following area of operations to mitigate fuel damage.
On April 13, 2011, while performing procedure reviews as part of industry-wide
self-assessments in response to the core damage events at Fukushima Daiichi,
Wolf Creek engineers identified two instances of mitigating strategy procedures
which did not contain sufficient information to accomplish those strategies
successfully. The first example was the ability to refill the refueling water storage
tank, and the second example involved flashing the diesel generator field using
alternate dc sources. These issues were documented in the licensees corrective
action program as Condition Report 37374. The inspectors evaluated these
findings under Inspection Manual Chapter 0609, Appendix L, and determined
these findings to be of very low safety significance because the findings did not
involve unrecoverable unavailability of multiple mitigating strategies such that
spent fuel pool cooling, injection to the reactor vessel, or injection to steam
generators cannot occur, or unrecoverable unavailability of on-site, self-powered,
portable pumping capability, or substantial inability to perform command and
control enhancements.
.2 Title 10 CFR Part 50, Appendix B, Criterion XI, Test Control, requires, in part,
that a test program be established to assure that all testing required to
demonstrate that structures, systems and components will perform satisfactorily
in service is identified and performed in accordance with written test procedures
which incorporate the requirements and acceptance limits contained in the
applicable design documents. On May 13, 2011, Wolf Creek identified a
noncited violation of 10 CFR Part 50, Appendix B, Criterion XI, test control for
stroking residual heat removal containment sump valve 8811B prior to its as-
found diagnostic test. Wolf Creek stroked the valve for a clearance order and as
such, preconditioned the valve prior to its test. Plant computer data from this
stroke, data from the diagnostic stroke, and valve disassembly showed no
deficiencies. Using Inspection Manual Chapter 0609.04, the inspectors
determined the finding to be of very low safety significance because it was
confirmed not to result in the loss of operability or functionality. This issue is
captured in Condition Report 37244.
- 68 - Enclosure
SUPPLEMENTAL INFORMATION
KEY POINTS OF CONTACT
Licensee Personnel
G. Beckett, Superintendent, Support Engineering
P. Bedgood, Manager, Radiation Protection
R. Evenson, Requalification Program Supervisor
J. Harris, System Engineer
S. Hedges, Site Vice President
S. Henry, Operations Manager
R. Hobby, Licensing Engineer
D. Hooper, Supervisor, Regulatory Affairs
T. Just, Senior Technician, Chemistry
J. Keim, Support Engineering Supervisor
S. Koenig, Manager, Corrective Actions
M. McMullen, Technician, Engineering
C. Medency, Supervisor, Radiation Protection
W. Muilenburg, Licensing Engineer
R. Murray, Simulator Supervisor
B. Norton, Manager, Integrated Plant Scheduling
J. Pankaskie, Engineering Supervisor
G. Pendergrass, Director of Engineering
L. Rockers, Licensing Engineer
G. Sen, Regulatory Affairs Manager
R. Smith, Plant Manager
M. Sunseri, President and Chief Executive Officer
J. Truelove, Supervisor, Chemistry
J. Weeks, System Engineer
M. Westman, Training Manager
NRC Personnel
D. Loveless, Senior Reactor Analyst
LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED
Opened
05000482/2011003-07 VIO Failure to Correct Procedure for Opening Main Steam Isolation
Valves (EA-11-149) (Section 4OA3.1)
A-1 Attachment
Opened and Closed
05000485/2011003-01 NCV No Procedure for Debris in Transformed and Tank Yards Prior to
Severe Weather (Section 1R01)05000482/2011003-02 NCV Failure to Properly Establish Clearance Order Boundary Isolation
Resulting in Loss of Component Cooling Water Inventory
(Section 1R04)05000482/2011003-03 NCV Failure to Assure Fillet Weld Met Size Requirements on Train B
Charging Header Vent Line (Section 1R08.1)05000482/2011003-04 NCV Failure to Assure Separation of Stainless Steel and Carbon Steel
Grinding and Cutting Equipment (Section 1R08.1)05000482/2011003-05 NCV Failure to Assure Configuration Control of Safety-Related
Systems (Section 1R08.3)05000482/2011003-06 FIN Inadequate Acceptance Criteria for Postmaintenance Testing of
the Startup Feedwater Pump (Section 1R19)05000482/2011003-08 NCV Failure to Maintain Reactor Coolant System Pressure Below
Relief Valve Setpoint (Section 4OA3.2)05000482/2011003-09 NCV Inadequate Fire Watch Defeats Halon Fire Suppression in Vital
Switchgear Rooms During Fire (Section 4OA3.3)05000482/2011003-10 NCV Failure to Analyze for Vortexing in Containment Spray Additive
Tank (Section 4OA5.1)
05000482-2515/177 TI Managing Gas Accumulation in Emergency Core Cooling, Decay
Heat Removal, and Containment Spray Systems (NRC Generic
Letter 2008-01) (Section 4OA5.1)
Closed
05000482/2006-003-00 LER Indications Discovered on Pressurizer during Preplanned In-
service Inspections (Section 4OA3.4)05000482/2010006-05 VIO Notice Of Violation EA-10-160, Failure to correct NRC identified
NCV. Apparent Cause Evaluation vice Root Cause Evaluation for
Essential Service Water (Section 4OA3.5)
A-2 Attachment
05000482-2515/183 TI Followup to the Fukushima Daiichi Nuclear Station Fuel Damage
Event (Section 4OA5.2)
05000482-2515/184 TI NRC Temporary Instruction 2515/184, Availability and Readiness
Inspection of Severe Accident Management Guidelines (SAMGs)
(Section 4OA5.3)
05000482/2515/172 TI Temporary Instruction 2515/172, Reactor Coolant System
Dissimilar Metal Butt Welds (Section 4OA5.4)
LIST OF DOCUMENTS REVIEWED
Section 1R01: Adverse Weather Protection
PROCEDURES
NUMBER TITLE REVISION
OFN AF-025 Unit Limitations 32
Ai 14-008 Severe Weather 9A
AP 21C-001 Wolf Creek Substation 11A
OP1450001 Outage Risk Management 000
APF 22B-001-05 Shutdown Risk Assessment 0
APF 22B-001-10 Shutdown Safety function Status and Assessment Summary 1
CONDITION REPORTS
00040573 00040351
WORK ORDERS
11-344384-000
A-3 Attachment
Section 1R04: Equipment Alignment
PROCEDURES
NUMBER TITLE REVISION /
DATE
CKL EN-120 Containment Spray System Lineup 15A
AP 21E-001 Clearance Orders 27
SYS EN-400 Containment Spray System Fill and Vent 11
STN EN-003A Containment Spray Train A Void Monitoring and Venting 3
STN EN-003B Containment Spray Train B Void Monitoring and Venting 3
CKL HB-122 Liquid Waste Evaporator Normal Lineup 15
D-HB-N-029 Clearance Order Liquid Radwaste System March 30,
2011
Standing Order 1 Valve Setup and Operation 43
M-12HB01 Piping and Instrumentation Diagram Liquid Radwaste 19
System
CONDITION REPORTS
13599 25918 28343 28771 32378
33060 33063 33064 34505
WORK ORDERS
93-100775-001 94-100830-001 03-257175-003
DRAWINGS
NUMBER TITLE REVISION
M-13EN03 Piping Orthographic Containment Spray System Reactor 3
Building A & B Trains
M-13EN05 Piping Orthographic Containment Spray System Reactor 2
Building A & B Trains
A-4 Attachment
M-12EN01 Piping and Instrumentation Diagram Containment Spray 12
System
M-13EN01 Piping Isometric Containment Spray System Auxiliary 7
Building A Train
M-13EN01 Piping Isometric Containment Spray System Auxiliary 7
Building B Train
M-13EN06 Small Piping Isometric Containment Spray System Auxiliary 0
Building
Section 1R05: Fire Protection
PROCEDURES
NUMBER TITLE REVISION
E-1F9905 Fire Hazard Analysis 0
AP 10-106 Fire Preplans 7
FPPM-001 Auxiliary Bldg El. 1974 2
DRAWINGS
NUMBER TITLE REVISION
M-663-00017 Penetration Typical Details, Attachment B W21
AP 10-106 Fire Preplans 7
FPPM-001 Auxiliary Bldg El. 1974 2
A-5 Attachment
CONDITION REPORTS
15073
Section 1R08: Inservice Inspection Activities
PROCEDURES
NUMBER TITLE REVISION
AI 16F-001 Evaluation Of Boric Acid Leakage 5A
AI 16F-002 Boric Acid Leakage Management 7
AI 28A-010 Screening Condition Reports 8A
AP 16F-001 Boric Acid Corrosion Control Program 5A/6A
AP 28-100 Condition Reports 13
AP 29A-003 Steam Generator Management 14
APF 28D-001 Self-Assessment Process 12
I-ENG-023 Steam Generator Data Analysis Guidelines 11
MRS 2.4.2 GEN-35 Eddy Current Inspection of Preservice and Inservice 14
Heat Exchanger Tubing
PDI-ISI-254-SE-NB Remote Inservice Examination of Reactor Vessel 1
Nozzle to Safe End, Nozzle to Pipe, and Safe End
to Pipe Welds Using the Nozzle Scanner
Generic Procedure for the Ultrasonic Testing of E
PDI-UT-2
Austenitic Pipe Welds
Generic Procedure for Ultrasonic Through Wall D
PDI-UT-3
Sizing in Pipe Welds
Generic Procedure for the Ultrasonic Testing of F
PDI-UT-6
Generic Procedure for the Ultrasonic Examination of F
PDI-UT-8
Weld Overlaid Similar and Dissimilar Metal Welds
QCP-20-501 PT (Penetrant Testing) 8
A-6 Attachment
QCP-20-502 Magnetic Particle Examination AC/DC Yoke and AC 8B
Coil Techniques
QCP-20-503 UT Thickness-Wall Thin 3
QCP-20-504 UT For Flaw Detection 5
QCP-20-508 Radiographic Examination of Welds 4A
QCP-20-510 UT Instrument Linearity 3
QCP-20-511 RT of AWS Groove Welds 1B
QCP-20-514 ET Testing 5B
QCP-20-516 PT/NON-STD Temp 05
QCP-20-517 RT Wall Thinning 2A
QCP-20-520 Pressure Test Examination 8B
QCP-20-521 UT Profile and Plotting 1B
QCP-20-522 UT Ferritic Pipe Welds 1B
QCP-20-523 UT Austenitic Pipe Welds 1B
QCP-20-527 UT- Soldering 1
QCP-20-540 VT-1 Visual Exam 0C
QCP-20-541 VT-3 Visual Exam 2A
QCP-20-543 Fluorescent Dye PT Exam 1
QCP-20-600 Visual Examination Of ASME Welds 9A
SG-CDME-10-8 Wolf Creek Steam Generator Secondary Side 0
Condition Monitoring Assessment and Operational
Assessment For Fuel Cycle and Refueling
Outage 18, February 2011
Wolf Creek, RF18 Condition Monitoring Assessment 2
SG-SGMP-09-23
and Operational Assessment, November 2009
Wolf Creek, RF18 Steam Generator Degradation 1
SG-SGMP-10-30
Assessment, March 2011
A-7 Attachment
Foreign Object Search and Retrieval and 11
STN PE-370
Secondary Side Inspections
STN PE-040D RCS Pressure Boundary Integrity Walkdown 3
STS PE-022 Steam Generator Tube Inspection 18
STS PE-040E RPV HEAD VISUAL INSPECTION 2
Ultrasonic Examination of Vessel Welds and 28
UT-2
Adjacent Base Metal
UT-95 Ultrasonic Examination of Austenitic Piping Welds 3
WCRE-24 WESDYNE Year 2011 Reactor Vessel Nozzle Safe- 0
end Examinations Program Plan
WCAL-002 Pulser/Receiver Linearity Procedure 10
WDI-CAL-102 Calibration Procedure for PCI Eddy Current Card 1
Installation and Removal of the WESDYNE Nozzle 5
WDI-EQPT-1021
Scanner (SQUID)
WDI-EQPT-1022 Reactor Vessel Nozzle Scanner Setup and 4
Checkout
WDI-STD-146 ET Examination of Reactor Vessel Pipe Welds 11
Inside Surface
CONDITION REPORTS
21975 28474 21976 28601 22027 28771
22128 28847 22280 28848 22391 28959
23173 28967 23251 28978 23455 29128
23459 29197 23867 29237 24020 29612
24077 29801 24230 30023 24336 30067
24339 30210 24469 30899 24658 31003
24659 31366 24661 31742 24662 31763
24663 31765 24665 31766 24676 31779
24681 31799 24857 31808 24893 31865
25095 32035 25173 32115 25196 32117
25224 32203 25228 32204 25268 32298
25361 32559 25377 32412 25394 32646
25495 32648 25643 32842 25871 33225
26354 33355 26358 33575 27193 33581
A-8 Attachment
27472 33600 27650 33603 27892 33684
28050 33686 28144 33688 28258 33690
28322 33689 28386 35793 28403
WORK ORDERS
08-310289 10-326485 10-324683 10-326486 10-325740
10-324621 09-320607 10-325747 10-325738 10-326483
10-325742
MISCELLANEOUS
NUMBER TITLE REVISION /
DATE
2010 3rd Quarter Outside Containment BACCP
Monitoring Walk-down
Boric Acid Leakage Screening/Evaluation for Normal Train May 5, 2010
B Charging Pump (PBG04)
Boric Acid Leakage Screening/Evaluation for Reactor March 5, 2010
Coolant Pump A (PBB01A)
Boric Acid Leakage Screening/Evaluation for Accumulator October 19,
Tank C Discharge Check Valve (EP8956C) 2009
Boric Acid Leakage Screening/Evaluation for RHR HX October 20,
A/CVCS To SI Pump A Upstream Isolation (EMHV8924) 2009
Boric Acid Leakage Screening/Evaluation for SI Pump B July 8, 2010
Suction Check Valve (EM8926B)
Boric Acid Leakage Screening/Evaluation for SI Pump A September 8,
Suction Check Valve (EM8926A) 2010
Boric Acid Leakage Screening/Evaluation for RCS Loop 3 November 20,
Steam Generator Primary Side Downstream Drain 2009
(BBV0476)
Boric Acid Leakage Screening/Evaluation for RCS Loop 1 March 5, 2010
Steam Generator Primary Side Downstream Drain
(BBV0474)
A-9 Attachment
Change Package # 012869, Installation of Vent Valves in 3
the Chemical and Volume Control System (BG), the
Residual Heat Removal (EJ), and the High-Pressure
Coolant Injection System (EM)
Technical Report No. 11-2039-TR-001, Failure Analysis of March 2011
Socket Weld on a Vent Valve Assembly from the CVCS
S/G Eddy Current Calibrated Equipment List October 16,
2009
Steam Generator data Analysis Desktop Instruction 4
RF 18 Steam Generator data Analysis Desktop Instruction 0
SGAMP Self Assessment, Steam Generator Asset October 17,
Management Program 2008
Wolf Creek RF 17 Fall 2009 Steam Generator Secondary August 17,
Side Visual Inspection Recommendations 2009
Wolf Creek RF17 Condition Monitoring Assessment and November
Operational Assessment 2009
APF 28D-001-02 Self Assessment Report SEL 04-038 , Steam Generator 4
Program
APF 29A-003-001 Secondary Chemistry Wet Layup Initial Monitoring 2
Frequency
ET 09-0016 Revision to Technical Specifications 5.5.9, "Steam June 2, 2009
Generator (SG) Program," and TS 5.6.10, "Steam
Generator Tube Inspection Report, for a Permanent
Alternate Repair Criterion
ET-09-0025 Docket No. 50-482: Revision to Technical Specification September 15,
(TS) 5.5.9, "Steam Generator (SG) Program," and 2009
TS 5.6.10, "Steam Generator Tube Inspection Report"
ET-10-0030 Revision to Technical Specifications 5.5.9, Steam November 30,
Generator (SG) Program, and TS (Technical 2010
Specifications) 5.6.10, Steam Generator Tube Inspection
Report, for a Temporary Alternate Repair Criterion
A-10 Attachment
I3R-01 Wolf Creek Generating Station - Third 10-Year Interval February 21,
Inservice Inspection Program Relief Request I3R-01 (TAC 2007
NO. MD0297)
I3R-05 Wolf Creek Generating Station - Authorization Of Relief July 19, 2007
Request I3R-05, Alternatives To Structural Weld Overlay
Requirements (TAC NO. MD1813)
13R-06 Wolf Creek Generating Station -Relief Request 13R-06, July 23, 2009
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)
Docket 10 CFR 50.55a Request 13R-06, Alternative to the September 16,
No. 50-482 Examination Requirements of ASME Section XI for 2008
Class 1 Piping Welds Examined from the Inside of the
Reactor Vessel
Docket Wolf Creek Nuclear Operating Corporation's Response to April 23 , 2009
No. 50-482 Request for Additional Information Regarding
10 CFR 50.55a Request 13R-06
ET 05-0014 10 CFR 50.55a Request Number 13R-03 for the Third September 28
Ten-Year Interval lnservice lnspection (ISI) Program - 2005
Request for Relief to Allow Use of Alternate Requirements
for Snubber lnspection and Testing
ET 06-0042 Wolf Creek Nuclear Operating Corporation's Response to September 27,
the September 20, 2006 NRC Request for Additional 2006
Information Regarding 10 CFR 50.55a Request 13R-05
ET 06-0044 Wolf Creek Nuclear Operating Corporations Revised October 2,
Commitment Regarding 10 CFR 50.55a Request 13R-05 2006
ET 06-001 0 Inservice Inspection Program Plan for the Third Ten-Year March 2, 2006
Interval and 10 CFR 50.55a Requests 13R-01, 13R-02,
and 13R-04
ET 06-0021 10 CFR 50.55a Request 13R-05, Installation and May 19, 2006
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
A-11 Attachment
ET 06-0031 Wolf Creek Nuclear Operating Corporation's Response to August 4,2006
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 06-0043 Wolf Creek Nuclear Operating Corporation's Response to October 5,200
NRC Request for Additional Information Regarding 10 6
CFR 50.55a Request 13R-01
ET 06-0058 Wolf Creek Nuclear Operating Corporation's Response to December 20,
the Second NRC Request for Additional Information 2006
Regarding 10 CFR 50.55a Request 13R-01
MRS-TRC-2087 Use of Appendix H and I Qualified Techniques at Wolf 0
Creek RF18 April 2011 S/G Inspection
SAP-+PT-09 Steam Generator Eddy Current Inspection Multi- 0
Frequency Eddy Current Parameters
SAP-+PTUB-09 Steam Generator Eddy Current Inspection Multi- 0
Frequency Eddy Current Parameters
SAP-01-09 Steam Generator Eddy Current Inspection Multi- 0
Frequency Eddy Current Parameters
SAP-02-09 Steam Generator Eddy Current Inspection Multi- 0
Frequency Eddy Current Parameters
SAP-03-09 Steam Generator Eddy Current Inspection Multi- 0
Frequency Eddy Current Parameters
SAP-04-09 Steam Generator Eddy Current Inspection Multi- 0
Frequency Eddy Current Parameters
SAP-05-09 Steam Generator Eddy Current Inspection Multi- 0
Frequency Eddy Current Parameters
SAP-06-09 Steam Generator Eddy Current Inspection Multi- 0
Frequency Eddy Current Parameters
SAP-07-09 Steam Generator Eddy Current Inspection Multi- 0
Frequency Eddy Current Parameters
SAP-08-09 Steam Generator Eddy Current Inspection Multi- 0
Frequency Eddy Current Parameters
SAP-09-09 Steam Generator Eddy Current Inspection Multi- 0
Frequency Eddy Current Parameters
A-12 Attachment
SAP-10-09 Steam Generator Eddy Current Inspection Multi- 0
Frequency Eddy Current Parameters
SAP-11-09 Steam Generator Eddy Current Inspection Multi- 0
Frequency Eddy Current Parameters
SAP-12-09 Steam Generator Eddy Current Inspection Multi- 0
Frequency Eddy Current Parameters
SAP-BOB-09 Steam Generator Eddy Current Inspection Multi- 0
Frequency Eddy Current Parameters
SAP-DELTA-09 Steam Generator Eddy Current Inspection Multi- 0
Frequency Eddy Current Parameters
SAP-GHENT-09 Steam Generator Eddy Current Inspection Multi- 0
Frequency Eddy Current Parameters
SEL 04-038 Steam Generator Program 4
SG-CDME-08-15 Wolf Creek RF16 Condition Monitoring Assessment and 1
Operational Assessment, April 2008
SG-CDME-09-1 Wolf Creek Steam Generator Secondary Side Condition 0
Monitoring and Operational Assessment for Fuel Cycle
and Refueling Outage 17
SG-SGMP-09-9 Steam Generator Degradation Assessment for Wolf 0
Creek, RF17 Refueling Outage, October 2009
SEL 09-151 EPRI-WRTC/Utility Welding Program Best Practices
Visual Examination for Leakage of PWR Reactor Head 2
WCRE-15 Program Plan For Management Of Alloy 600 Components 3
And Alloy 82/182 Welds
Section 1R11: Licensed Operator Requalification Program
NUMBER TITLE REVISION
LR5002026 Inadvertent Safety Injection Lab 3
A-13 Attachment
Section 1R12: Maintenance Effectiveness
MISCELLANEOUS
NUMBER TITLE
GK-01 Final Scope Evaluation, System GK, Control Building HVAC
System
CONDITION REPORTS
00035992 00027105 00026250 00028792 00027228
00026251 00034299 00026250
Section 1R13: Maintenance Risk Assessment and Emergent Work Controls
MISCELLANEOUS
NUMBER TITLE REVISION
EDI 23M-050 Engineering Desktop Instruction Monitoring Performance to 3
Criteria and Goals
MPE GK-003 Control Room and Class 1E A/C Units Preventive 3
Maintenance Activity
EDI 23M-050 Engineering Desktop Instruction Monitoring Performance to 3
Criteria and Goals
WORK ORDERS
10-330270-000 10-330269-000 10-330270-000 10-330269-000
Section 1R15: Operability Evaluations
NUMBER TITLE REVISION
ALR 501 Standby Diesel Engine System Control Panel KJ-121 13, 14 and 14A
A-14 Attachment
AP 26C-004 Operability Determination and Functionality Assessment 23
OE KJ-10-001 Emergency Diesel Generators KKJ01A and KKJ01B 0
MISCELLANEOUS
NUMBER TITLE REVISION /
DATE
EDI 23M-050 Engineering Desktop Instruction Monitoring Performance to 3
Criteria and Goals
Final Scope Evaluation, System GK, Control Building HVAC
System
MPE GK-003 Control Room and Class 1E A/C Units Preventive 3
Maintenance Activity
EDI 23M-050 Engineering Desktop Instruction Monitoring Performance to 3
Criteria and Goals
GK Final Scope Evaluation - Control Building HVAC System
LER 22011-003-00, Diesel Generator Declared Inoperable May 12, 2011
Due to Inadequate Reinstallation of Pipe connection
Resulting in Excessive Governor Oil Coolant Leak
A-EDG Governor Heat Exchanger Water Leak
9.5-16 USAR 19
2011-1027-0 Training Needs Analysis
Operations Requalification Cycle 11-01 Week 0 to Week 6
Schedule
OP1336001 Plant Changes 0
A-15 Attachment
M-018-00110-W13 Electrical Schematic Engine Guide Panel KJ121
E-13KJ02 Schematic Diagram Diesel Generator KKJ01A Annunciator 7
and Miscellaneous Circuits
M-12EF01 Piping & Instrumentation Diagram Essential Service Water 57
System
M-12EF02 Piping & Instrumentation Diagram Essential Service Water 26
System
M-K2EF03 Piping & Instrumentation Diagram Essential Service Water 10
System
M-13EF07(Q) Piping Isometric Essential Service WTR.Sys.Control Bldg 1
Cooler(A&B) Train Supply & Return
M-13EF08 Piping Isometric Essential Service Wtr,-Diesel Generator 01
Bldg.
CONDITION REPORTS
00034661 00038229
REPORTABILITY EVALUATION REPORT
2011-037
Section 1R18: Plant Modifications
NUMBER TITLE REVISION
10-017-EG Temporary Cooling of CCW Radwaste Loads 0
CONDITION REPORTS
00035262
A-16 Attachment
Section 1R19: Postmaintenance Testing
PROCEDURES
NUMBER TITLE REVISION
SYS EG-205 CCW Flow Adjustment to Reactor Coolant Pumps, Seal 9
Water Heat Exchanger, and Excess Letdown Heat
Exchanger
STS AE-209 Main Feed Reg Valve Bypass Valve Inservice Valve Test 2
STN AE-001 Startup Main Feedwater Pump Operational Test 0A
STN AC-007 Turbine Overspeed Trip Set 26
AP 16E-002 Post Maintenance Testing Development 9C
WORK ORDERS
37698 38443 29128 34806 39145
34434 34500 36164
MISCELLANEOUS DOCUMENTS
NUMBER TITLE REVISION /
DATE
Control Room Turnover Checklist-Day Shift March 16,
2011
APF 30E-004-01 Main Feedwater System 2
BD-EMG FR-h1 Response to Loss of Secondary Heat Sink 10
FWIS and Reactor Trip on Low S/G LevelCR 29128 Root
Cause Evaluation
A-17 Attachment
E-0099 Cable Sheath Grounding and Termination Data 7
KD-7496 One Line Diagram 40
CONDITION REPORTS
0037698 00025817 00038443
WORK ORDERS
11-337610-000 11-337610-001 11-337610-002 11-337610-003 11-337610-004
11-337610-005 11-337610-006 11-337610-007 09-322525-000 10-335457-001
11-337610-004 11-337610-005 11-337610-006 11-337610-007 11-337610-000
11-337610-001 11-337610-002 11-337610-003
Section 1R20: Refueling and Other Outage Activities
PROCEDURES
NUMBER TITLE REVISION
FHP 02-007A Reactor Vessel Closure head Removal 10
SYS BB-215 RCS Drain Down with Fuel in Reactor 28
STS IC-439 Channel Calibration NIS Post Accident Monitoring N60 3A
GEN 00-002 Cold Shutdown to Hot Standby 73
STN EJ-002 Containment Inspection 17
DRAWINGS
NUMBER TITLE REVISION
C-OL2901(Q) Reactor Building Line Plate Floor Details, SHT-1 7
A-18 Attachment
C-OS2919(Q) Reactor Building Incore Instrumentation Tube Supports and 8
Platforms
C-OL2914(Q) Reactor Building Liner Place Floor Details-Sheet-3 4
MISCELLANEOUS DOCUMENTS
NUMBER TITLE REVISION /
DATE
11-2039-L-001 ALTRAN SOLUTIONS April 13, 2011
Refuel 18, No. 15 The Daily Howl April 2, 2011
Refuel 18, No. 17 The Daily Howl April 4, 2011
Information Failure of Motor Operated Value Actuator Motors with December 8,
Notice 2008-20 Magnesium Allow Rotors 2008
MS-02 Piping Class Sheets 53
Evaluation of Interim Operation 0
ALARA Planning Survey
RF18 High Impact Teams/Major Projects
Letter NE 11-0009, dated February 28, 2011, from W. H.
Ketchum To R. A. Smith and R. E. Kopecky
AP 22B-001 Outage Risk Management 13
CONDITION REPORTS
00029149 00029322 00030371 00030371 00032254
00033358 00033698 00033699 00033715 00033716
00034068 00034349 00035261 00035262 00035304
00035314 00035419 00035426 00035516 00035533
A-19 Attachment
00035535 00035537 00035539 00035540 00035540
00035541 00035542 00035544 00035545 00035546
00035547 00035548 00035549 00035550 00035551
00035552 00035553 00035554 00035555 00035556
00035557 00035558 00035559 00035560 00035573
00035614 00035615 00035617 00035619 00035620
00035621 00035622 00035623 00035624 00035625
00035626 00035627 00035628 00035629 00035630
00035632 00035663 00035714 00035963 00035965
00035987 00036031 00036032 00036106 00036186
00036272 00036292 00036300 00036492 00036518
00036798 00036799 00036857 00036876 00036880
00036881 00036957 00036966 00036988 00037110
00037289 00037615 00037909 00038083 00038086
00038113 00038333 00038517 00038680 00039099
00039283 2007-000299 00035429 00039721
REPORTABILITY EVALUATION REQUESTS
2011-047 2011-102 2011-103 2011-113 00037048
2011-059 2011-057 2011-046 2011-056
WORK ORDERS
10-324324-000 10-328967-001 10-324322-000 10-339212-003
Section 1R22: Surveillance Testing
PROCEDURES
NUMBER TITLE REVISION
STS BG-002A Train A ECCS System Vent for Mode 4 10
STS BG-007A ECCS Valve Check and Train A and Common Void 5
Monitoring and Venting
A-20 Attachment
Section 1R22: Surveillance Testing
PROCEDURES
NUMBER TITLE REVISION
STS BG-007B ECCS Train B Void Monitoring and Venting 5
SYS EM-410 Fill and Vent of Safety Injection System After Maintenance 17 and 18
SYS EJ-110 RHR System Fill and Vent Including Initial RCS Fill 56, 57, 59
and 60
STS IC-211B Actuation Logic Test Train B Solid State Protection System 35
STS PE-018 Containment Integrated Leakage Rate Test 9
AP 21-004 Operator Response Time Program 2
STN TCA-001 Manual Time Critical Action Timing 3
SYS GP-519 CILRT-EN System 2
AP 29E-001 Program Plant for Containment Leakage Measurement 13
AI 21-016 Operator Time Critical Actions Validation 2
STS KJ-001B Integrated Diesel Generator and Safeguards Actuation 42A
Testing Train B
STS IC-615B Slave Relay Test K615 Train B Safety Injection 27
WORK ORDERS
10-326512-001 09-322158-001
CONDITION REPORTS
00037110 00037244 30302 33443 39083
A-21 Attachment
39081
REPORTABILITY EVALUATION REQUESTS
2011-094
MISCELLANEOUS DOCUMENTS
NUMBER TITLE REVISION
AN-99-025 Steam Generator Tube Rupture Overfill Analysis with 1
Revised Operator Action Times
AIF 21-016-02 Time Verification Form 0A
EJHV8811B Analysis Print
EJHV8811B Refuel XVIII Preparation Package
Section 4OA2: Identification and Resolution of Problems
PROCEDURES
NUMBER TITLE REVISION
KMS-4 Mechanical Standard 2
VENDOR DOCUMENTS
NUMBER TITLE REVISION
TB-68-2 Tensile Strength of Threaded Insert Assembly 2
CONDITION REPORTS
38321
WORK ORDERS
11-339714-000 11-337546-000
A-22 Attachment
Section 4OA3: Event Follow-Up
PROCEDURES
NUMBER TITLE REVISION
BD-EMG ES-03 SI Termination 10A
SYS AE-200 Feedwater Preheating During Plant Startup and Shutdown 29 and 30
AP 15C-002 Procedure Use and Adherence33
EMG E-0 Reactor Trip or Safety Injection 25
SYS AB-120 Main Steam and Steam Dump Startup and Operation 27
SYS PN-200 Energizing and Deenergizing Inverters PN09 and PN10 11
ALR KC-888 Fire Protection Panel KC-008 Alarm Response 18A
AP 10-104 Breach Authorization 24A
SYS GK-200 Inoperable Class IE A/C Unit 21A
AP-10-103 Fire Protection Impairment Control 23A
AP 21D-003 Control of Information Tagging 15B
SYS BG-120 Chemical and Volume Control System 42
GEN 00-006 Hot Standby to Cold Shutdown 76
AP 21-001 Conduct of Operation 50
SY1300400 Chemical and Volume Control System - Low Pressure 13
Letdown
SY1300400 Chemical and Volume Control System - Plan/Text 25
A-23 Attachment
MISCELLANEOUS DOCUMENTS
NUMBER TITLE REVISION /
DATE
CR 34964 White Streamline Safety Injection When ABHV20 Was Opened 2
Paper
Change Package PRT Sparing Line Bypass 0
012410
Change Package CRDM Nozzle #6 Scratch Evaluation 0
013674
ES-1.1 Background Information for Westinghouse Owners Group April 30, 2005
Emergency Response Guideline
Control Room Turnover Checklist March 11, 19
and 22, 2011
Corrective Action Review Board Meeting Minutes March 23,
2011
Chapter 7.3-39 Updated Safety Analysis Report 13
Site Clock Reset Communication - Condition Report 34964 March 19,
2011
Page 14 of 17 Breach Authorization Permit Log April 22, 2011
Page 6 of 7 Fire Protection Impairment Control Log April 22, 2011
2011-118, 119, Fire Protection Impairment Control Permit
121, 122
2011-148, 149, Breach Authorization Permit
215, 237, 238
Fire Protection Significance Determination Review
04/05/2011 Halon Discharge in ESF Switchgear Room 1
Fire Incident Investigation Report April 5, 2011
FW1431401 Just-in-Time Training - Alternate Planning and/or Training 0
Record
A-24 Attachment
11-339929-001 AMETEK Solidstate Controls
16577-M-658 Technical specification for Furnishing, Installing, and Testing
Halogenated Agent Extinguishing System for the
Standardized Nuclear Unit Power Plant System (SNUPPS)
Wolf Creek Unit Only
SU4-KC02 Fire Protection System Halon Preoperational Test 0
Control Room Turnover Checklist April14, 15,
21, 2011
Boundary Watch Duties
FW1231401 Fire Watch Duties and Responsibilities 10
LR5005012 JIT Plant Shutdown From 100% RTP 3
I-11154 Operation and Maintenance Instructions Solenoid Power 1
Operated Relief Valve
59 99-0007 LTOP TS Bases - B 3.4.12 1
DRAWINGS
NUMBER TITLE REVISION
M-744-00019 SNUPPS Projects Functional Diagram Reactor Trip Signals W07
M-744-00024 SNUPPS Projects Functional Diagram Steam Generator Trip W06
Signals
M-744-00025 SNUPPS Projects Functional Diagram Safeguards Actuation W07
Signals
CONDITION REPORTS
00033745 00034963 00034964 00034964 00034967
00034968 00034969 00034970 00034975 00034987
00034995 00035000 00035001 00035012 00035017
00035246 00035249 00035251 00035319 00035333
A-25 Attachment
00035515 00035638 00035648 00035650 00035652
00036164 00036719 00037931 00038232 00038516
REPORTABILITY EVALUATION REQUEST
2011-040
Work Orders
08-310440-001 08-310449-000 08-310449-001 08-310440-000 11-339200-001
11-339027-000
Section 4OA5: Other Activities
CALCULATIONS
NUMBER TITLE REVISION
CN-SEE-III- Evaluation of Suction Side Gas Void Volumes for Wolf Creek to 0
11-6 Address GL-2008-01
EN-M-024 Critical Submergence in Containment Spray Additive Tank 0
(TEN01) to Avoid Vortex
XX-M-074 Comparison of GOTHIC Gas Transport Calculations with 0
Westinghouse Test Data for Wolf Creek Emergency Core Cooling
System
XX-M-076 Startup Pressure Pulse Analysis for WCGS ECCS Discharge 0
Piping
XX-M-079 ECCS (Emergency Core Cooling System) Horizontal Line 1
Metrology (laser measurements) Data Evaluation,
CONDITION REPORTS
00006250 00008212 00018673 00029160 00033057
00032378 00033060 00033061 00033062 00033063
A-26 Attachment
00033065 00033070 00033071 00038714 00038715
2008-000091
DRAWINGS
NUMBER TITLE REVISION
M-12BG03 Piping and Instrumentation Diagram Chemical & Volume 47
Control System
M-12BN01 Piping and Instrumentation Diagram Borated Refueling 14
Water Storage System
M-12EJ01 Piping and Instrumentation Diagram Residual Heat Removal 46
System, sheet 1
M-12EM01 Piping and Instrumentation Diagram High Pressure Coolant 37
Injection System
M-12EM02 Piping and Instrumentation Diagram High Pressure Coolant 19
Injection System
M-12EP01 Piping and Instrumentation Diagram Accumulator Safety 08
Injection System
M-12EN01 Piping and Instrumentation Containment Spray System 12
M-13EJ01 Piping Isometric Residual Heat Removal System - Auxiliary 09
Building A Train
M-13EM01 Piping Isometric High Pressure Coolant Injection System - 16
Auxiliary Building
A-27 Attachment
INSPECTION REPORTS (05000482/
2008007 2009006 2009007 2010005 2010006
2010007
MISCELLANEOUS DOCUMENTS
NUMBER TITLE REVISION /
DATE
Brooks Metrology Report
Examples of Accumulator Level Alerts
List of discharge and suction vent valves for the
Generic Letter 2008-01 systems
Technical Specifications Surveillance
Requirement 3.5.2.3 Bases
Updated Final Safety Analysis Report, Section 6.3.2.2
Updated Final Safety Analysis Report Change
Request 2008-004 to section 6.3.2.2
2008-01 Managing Gas Accumulation in Emergency Core January 11,
Cooling, Decay Heat Removal, and Containment Spray 2008
Systems
2008-0624 Technical Specifications Document Revision Request
APC 09-20 Generic Letter (GL) 2008-01, Managing Gas May 18, 2009
Accumulation in Emergency Core Cooling, Decay Heat
Removal, and Containment Spray Systems -
Evaluation of Unexpected Voids or Gas Identified in
A-28 Attachment
Plant ECCS and Other Systems
Form Engineering Screening Form 17
APF 05-002-01
Gas Voiding Improvement Plan - Project Report 1
NEI 09-10 Guidelines for Effective Prevention and Management 1
of System Gas Accumulation
Reviewed the set of laser metrology isometric drawings
Report FAI/08-70 Gas Voids Pressure Pulsations Program 1
Report FAI/11-192 Void Acceptance Criteria for Wolf Creek Discharge March /2011
Piping Based on FAI/08-70 Methodology, Revision 1
SEL 2011-196 STARS Gas Team Self-Assessment January 20,
2011
Specification M-204 Technical Specification for Field Fabrication and 46
Installation of Piping and Pipe Supports to ASME
Section III for the Wolf Creek Generating Station
Standing Order 33 Accumulator Level Alert E-mail 0
WCAP-16631-NP Testing and Evaluation of Gas Transport to the Suction 0
of ECCS Pumps - Volume 1
WCAP-17276-P Investigation of Simplified Equation for Gas Transport January 2011
WCNOC122-PR-01 Study of Vent Requirements for Cooling Water 0
Systems
A-29 Attachment
System Walk Down Reports
PROCEDURES
NUMBER NUMBER REVISION
AI 23P-001 Gas Intrusion Program 0
AP 21E-001 Clearance Orders 27
QCP-20-526 Ultrasonic Measurement for Liquid Level Measurement 1
STN IC-252A Calibration of RHR Pump A Mini Flow Valve Control Switch 7A
STN IC-252B Calibration of RHR Pump B Mini Flow Valve Control Switch 8A
STS BG-002 ECCS Valve Check and System Vent 26
STS BG-002A Train A ECCS System Vent for Mode 4 5 and 10
STS BG-002B Train B ECCS System Vent for Mode 4 4 and 10
SYS BG-120 Chemical and Volume Control System Startup 43
SYS EG-400 Component Cooling Water System Fill and Vent 20A
SYS EM-410 Fill and Vent of Safety Injection System After Maintenance 18A
SYS EJ-110 RHR System Fill and Vent Including Initial RCS Fill 60
SYS SJ-002 Void Sampling Using a Sample/Purge Rig 1
A-30 Attachment
SURVEILLANCE TESTS
TITLE TITLE DATE
STS BG-007A ECCS Valve Check and Train A and Common Void March 3, 2011
Monitoring and Venting
STS BG-007B ECCS Train B Monitoring and Venting March 18, 2011
STS EG-003A CCW Train A Monitoring and Venting March 16. 2011
STS EG-003B CCW Train A Monitoring and Venting March 16, 2011
STS EN-003A Containment Spray Train A and Common Void March 2, 2011
Monitoring and Venting
STS EN-003B Containment Spray Train B Void Monitoring and Venting March 15, 2011
Section 4OA5: Other Activities
NUMBER TITLE REVISION
EPP 06-021 Training Programs 8
SAM SAG-01 Inject into the Steam Generators 1
SAM SAG-02 Depressurize the RCS 1
SAM SAG-04 Inject into Containment 1
A-31 Attachment
SAM SAG-05 Reduce Fission Product Releases 1
SAM SAG-06 Control Containment Conditions 1
SAM SAG-07 Reducing Containment Hydrogen 1
SAM SAG-08 Flood Containment 1
SAM SAEG-01 TSC Long Term Monitoring 2
SAM SAEG-02 SAMG Termination 1
SAM SACRG-02 SACRG for Transients after TSC is Functional 2
SAM SACRG-01 Severe Accident Control Room Guideline Initial Response 2
WOG Severe Accident Management Guidance 1
CONDITION REPORTS
18664 18398
A-32 Attachment