ML081430159
ML081430159 | |
Person / Time | |
---|---|
Site: | Wolf Creek |
Issue date: | 05/21/2008 |
From: | Vincent Gaddy NRC/RGN-IV/DRP/RPB-B |
To: | Muench R Wolf Creek |
References | |
EA-08-052 IR-08-002 | |
Download: ML081430159 (65) | |
See also: IR 05000482/2008002
Text
UNITED STATES
NUC LE AR RE G UL AT O RY C O M M I S S I O N
R E GI ON I V
612 EAST LAMAR BLVD , SU I TE 400
AR LI N GTON , TEXAS 76011-4125
May 21, 2008
Rick A. Muench, 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 05000482/2008002
Dear Muench:
On April 7, 2008, the U.S. Nuclear Regulatory Commission (NRC) completed an integrated
inspection at your Wolf Creek Generating Station. The enclosed report documents the
inspection results, which were discussed on April 11, 2008, with Mr. Stephen Hedges and other
members of your staff.
The inspection 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, six NRC identified and two self-revealing findings of very
low safety significance (Green) are documented in this report. All of these findings were
determined to involve violations of NRC requirements. Additionally, three licensee-identified
violations of very low safety significance is listed in this report. However, because of the very
low safety significance and because the findings were entered into your corrective action
program, the NRC is treating these violations as noncited violations consistent with Section VI.A
of the NRC Enforcement Policy.
If you contest these noncited violations, you should provide a response within 30 days of this
inspection report, with the basis for your denial, to the U.S. Nuclear Regulatory Commission,
ATTN.: Document Control Desk, Washington, DC 20555-0001; with copies to the Regional
Administrator Region IV; the Director, Office of Enforcement, U.S. Nuclear Regulatory
Commission, Washington, DC 20555-0001; and the NRC Resident Inspector at the Wolf Creek
Generating Station.
Wolf Creek Nuclear Operating Corp. -2-
In accordance with 10 CFR 2.390 of the NRCs Rules of Practice, a copy of this letter and its
enclosure will be made available electronically for public inspection in the NRC
Public Document Room or from the Publicly Available Records component of NRCs document
system (ADAMS), accessible from the NRC Web site at
http://www.nrc.gov/reading-rm/adams.html (the Public Electronic Reading Room).
Sincerely,
/RA/
Vincent G. Gaddy, Chief
Project Branch B
Division of Reactor Projects
Docket No. 50-482
License No. NPF-42
Enclosure: Inspection Report 05000482/2008002
w/Attachment: Supplemental Information
cc w/enclosure:
Vice President Operations/Plant Manager
Wolf Creek Nuclear Operating Corp.
P.O. Box 411
Burlington, KS 66839
Jay Silberg, Esq.
Pillsbury Winthrop Shaw Pittman LLP
2300 N Street, NW
Washington, DC 20037
Supervisor Licensing
Wolf Creek Nuclear Operating Corp.
P.O. Box 411
Burlington, KS 66839
Chief Engineer
Utilities Division
Kansas Corporation Commission
1500 SW Arrowhead Road
Topeka, KS 66604-4027
Office of the Governor
State of Kansas
Topeka, KS 66612
Wolf Creek Nuclear Operating Corp. -3-
Attorney General
120 S.W. 10th Avenue, 2nd Floor
Topeka, KS 66612-1597
County Clerk
Coffey County Courthouse
110 South 6th Street
Burlington, KS 66839-1798
Chief, Radiation and Asbestos
Control Section
Kansas Department of Health and
Environment
Bureau of Air and Radiation
1000 SW Jackson, Suite 310
Topeka, KS 66612-1366
Ronald L. McCabe, Chief
Technological Hazards Branch
National Preparedness Division
DHS/FEMA
9221 Ward Parkway
Suite 300
Kansas City, MO 64114-3372
Wolf Creek Nuclear Operating Corp. -4-
Electronic distribution by RIV:
Regional Administrator (Elmo.Collins@nrc.gov)
DRP Director (Dwight.Chamberlain@nrc.gov)
DRS Director (Roy.Caniano@nrc.gov)
DRS Deputy Director (Troy.Pruett@nrc.gov)
Senior Resident Inspector (Steve.Cochrum@nrc.gov)
SRI, Callaway (David.Dumbacher@nrc.gov)
Branch Chief, DRP/B (Vincent.Gaddy@nrc.gov)
Senior Project Engineer, DRP/B (Rick Deese@nrc.gov)
Team Leader, DRP/TSS (Chuck.Paulk@nrc.gov)
RITS Coordinator (Marisa.Herrera@nrc.gov)
Only inspection reports to the following:
J. Adams, OEDO RIV Coordinator (John.Adams@nrc.gov)
P. Lougheed, OEDO RIV Coordinator (Patricia.Lougheed@nrc.gov)
R. Kahler, NSIR/DPR/EDP (Robert.Kahler@nrc.gov)
ROPreports Resource
OEMail Resource
WC Site Secretary (Shirley.Allen@nrc.gov)
SUNSI Review Completed: __VGG__ ADAMS: ; Yes No Initials: __VGG__
- Publicly Available Non-Publicly Available Sensitive ; Non-Sensitive
R:\_REACTORS\_WC\2007\WC2008-002RP-SDC.wpd ADAMS ML081430159
SRI:DRP/B RI:DRP/B C:DRS/EB1 C:DRS/EB2
SDCochrum CMLong RBywater LJSmith
/RA/ /RA/ /RA/ /RA/ GAPick for
5/21/2008 5/21/2008 4/28/2008 4/30/2008
C:DRS/OB C:DRS/PSB ACES/SES C:DRP/B
RELantz MPShannon MMVasquez VGGaddy
/RA/ /RA/ /RA/ /RA/
4/29/2008 4/28/2008 5/13/2008 5/21/2008
OFFICIAL RECORD COPY T=Telephone E=E-mail F=Fax
U. S. NUCLEAR REGULATORY COMMISSION
REGION IV
Docket: 50-482
License: NPF-42
Report: 5000482/2008002
Licensee: Wolf Creek Operating Corporation
Facility: Wolf Creek Generating Station
Location: 1550 Oxen Lane NE
Burlington, Kansas
Dates: January 1 to April 7, 2008
Inspectors: S. D. Cochrum, Senior Resident Inspector
C. M. Long, Resident Inspector
G. A. Pick, Senior Reactor Inspector
D. L. Stearns, Health Physics Inspector
Approved by: V. G. Gaddy, Chief, Project Branch B
-1- Enclosure
TABLE OF CONTENTS
SUMMARY OF FINDINGS ......................................................................................................... 3
REPORT DETAILS..................................................................................................................... 9
REACTOR SAFETY ................................................................................................................... 9
1R01 Adverse Wather Protection (71111.01)..........................................................................9
1R04 Equipment Alignment (71111.04) ............................................................................. 10
1R05 Fire Protection (71111.05) ......................................................................................... 11
1R11 Licensed Operator Requalification Program (71111.11) ............................................ 13
1R12 Maintenance Effectiveness (71111.12)...................................................................... 14
1R13 Maintenance Risk Assessments and Emergent Work Control (71111.13) ................. 14
1R15 Operability Evaluations (71111.15)............................................................................ 17
1R19 Postmaintenance Testing (71111.19) ........................................................................ 19
1R20 Outage Activities (71111.20) ..................................................................................... 20
1R22 Surveillance Testing (71111.22) ............................................................................... 22
1EP6 Drill Evaluation (71114.06) ........................................................................................ 24
2OS1 Access Control to Radiologically Significant Areas (71121.01) ..24
2OS2 ALARA Planning and Controls (71121.02)................................................................. 28
OTHER ACTIVITIES ................................................................................................................ 29
4OA1 Performance Indicator Verification (71151)................................................................ 29
4OA2 Identification and Resolution of Problems (71152)..................................................... 32
4OA3 Event Followup .......................................................................................................... 34
4OA5 Other Activities .......................................................................................................... 44
4OA6 Meetings, Including Exit............................................................................................. 47
4OA7 Licensee-Identified Violations .................................................................................... 47
SUPPLEMENTAL INFORMATION ..........................................................................................A-1
KEY POINTS OF CONTACT ...................................................................................................A-1
ITEMS OPENED, CLOSED, AND DISCUSSED ......................................................................A-1
LIST OF DOCUMENTS REVIEWED .......................................................................................A-2
LIST OF ACRONYMS ...........................................................................................................A-16
-2- Enclosure
SUMMARY OF FINDINGS
IR 05000482/2008002; 1/01 - 4/07/08; Wolf Creek Generating Station; Fire Protection,
Maintenance Risk Assessments and Emergent Work Control, Access Control to Radiologically
Significant Areas, Event Followup and Other Activities.
This report covered a 3-month period of inspection by resident inspectors and regional
specialists. The inspection identified eight Green findings, all of which are noncited violations.
The significance of most findings is indicated by their color (Green, White, Yellow, or Red) using
Inspection Manual Chapter 0609, "Significance Determination Process." Findings for which the
significance determination process does not apply may be Green or be assigned a severity level
after NRC managements review. The NRCs 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 and Self-Revealing Findings
Cornerstone: Mitigating Systems
- Green. The inspectors identified a noncited violation of Technical
Specification 5.4.1.d for failure to implement fire protection impairment control
permit requirements and compensatory measures when operators received a
trouble alarm on a fire detector in the auxiliary building. On January 26, 2008,
operators discovered that Detector KC-104-XCH-ID-006 did not have a fire
protection impairment control permit. This detector was adjacent to Detector KC-
104-XSH-ID-007 which was already inoperable in Impairment 2008-020. The
licensees administrative procedure required fire detection in the area to be
declared inoperable if two adjacent detectors are inoperable. This condition
existed for approximately 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and would have required a compensatory
continuous fire watch for the period that both detectors were inoperable. The
residents identified that the control room turnover checklist contains a section for
listing the KC008 alarms; however, the two turnover checklists for the two shifts
following the initial alarm did not identify Detector KC-104-XCH-ID-006 as a
Detector KC-008 alarm.
The failure to implement fire protection impairment control permit requirements
and establish compensatory measures for the auxiliary building 2026-foot level
was considered a performance deficiency. The finding was more than minor
because it was associated with the Mitigating Systems Cornerstone attribute of
protection against external factors and affected the cornerstone objective of
ensuring the availability of systems that respond to initiating events to prevent
undesirable consequences. Specifically, this issue relates to the protection
against fire example of protection against external factors attribute because the
detectors were inoperable without ensuring compensatory measures were in
place. The finding was of very low safety significance because it involved
compensatory measures for the fixed fire protection system and was assigned a
low degradation rating since less than 10 percent of the fire detectors in the area
were disabled. The finding has crosscutting aspects in the area of human
performance associated with work practices because the licensee failed to apply
appropriate human error techniques such as self and peer-checking techniques
to avoid committing errors H.4(a) (Section 1R05).
-3- Enclosure
- Green. A noncited violation of Technical Specification 3.8.1.B.4 was identified
when the licensee performed elective maintenance in the switchyard and
removed equipment from service that was prohibited by Technical Specifications
while in an extended diesel generator outage.
The inspectors determined that the failure to implement requirements of
Technical Specification 3.8.1.B.4 was a performance deficiency. The finding was
more than minor because it is associated with the equipment performance
attribute for the mitigating systems cornerstone; and, it affected the cornerstone
objective to ensure the availability, reliability, and capability of systems that
respond to initiating events to prevent undesirable consequences (i.e. core
damage). The finding was determined to be of very low safety significance
because the issue resulted in the Train B offsite power being inoperable, but
capable of supplying the safety bus for greater than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. Additionally, the
cause of the finding has a human performance crosscutting aspect in the area
associated with work control. Specifically, Wolf Creek did not ensure
STS-IC-805B was appropriately coordinated within organizations to assure plant
and human performance during the extended emergency diesel generator
allowed outage time H.3(b) (Section 1R13).
- Green. The NRC identified a noncited violation of Technical Specification 5.4.1
for failure to follow the operability process on discovery of the centrifugal
charging Pump A room cooler leak. On February 13, 2008, at 2:20 p.m., the
control room was notified of a leak from the room cooler for the centrifugal
charging pump. At that time, it could not be established if the leak would cause a
loss of structural integrity of essential service water. Wolf Creek made no log
entries at 2:20 p.m. stating its basis for immediate operability. At 3:50 p.m., Wolf
Creek control room logs documented that centrifugal charging Pump A had a
room cooler leak and structural integrity cannot be verified. Subsequent entry
into Technical Specification 3.7.8 for the essential service water Pump A caused
emergency diesel Generator A to be inoperable. Technical Specification 3.8.1,
Condition I, states that with three alternating current sources inoperable (both
emergency diesel generators and an offsite source), Technical
Specification 3.0.3 shall be entered. Wolf Creek exited Technical
Specification 3.0.3 at 4:13 p.m. when the inlet and outlet valves to centrifugal
charging Pump As room cooler were closed. The inspectors could not locate
any justification produced by Wolf Creek for the room coolers operability after
2:20 p.m.
The inspectors determined that the failure to follow the operability process is a
performance deficiency. The inspectors determined that this finding was more
than minor because if left uncorrected, it could become a more serious problem if
the operability process is not correctly applied. The finding screened to Phase 2
because the finding represents an actual loss of safety function of a single train
of high head injection. A bounding risk of Green results from the Phase 2
presolved worksheets using an exposure time of less than 3 days for the
Centrifugal Charging Pump (CCP) A [Fails to Run]. The inspectors also
determined that the finding had a human performance crosscutting aspect in the
area associated with decision making because the licensee failed to use
conservative assumptions in its operability decision and apply a requirement to
-4- Enclosure
demonstrate that the room cooler is operable in order to proceed rather than a
requirement to demonstrate that it is inoperable H.1(b) (Section 4OA3.2(ii)).
- Green. The inspectors identified a noncited violation of 10 CFR Part 50
Appendix B Criterion XVI, Corrective Action, because Wolf Creek failed to take
timely corrective actions to prevent failure of the centrifugal charging pump A
room cooler which resulted in a Notice of Enforcement Discretion (EA-08-052).
The inspectors found that room Cooler SGL12A experienced leaks in
October 1999, May 2003, October 2003, August 2004, October 2006, and again
in February 2008. On March 14, 2007, Wolf Creek chose to delay SGL12As
replacement until Refueling Outage 16 due to the required length of time to
replace the cooler. On February 13, 2008, a circumferential flaw on an H-bend
was discovered in SGL12A preventing it from performing its safety function.
Inspectors reviewed corrective action Procedure AP 28A-100, Condition
Reports, Revision 3 and found that a loss of a train to perform its safety function
was considered a significant deficiency requiring corrective action to prevent
recurrence. The inspectors reviewed this issue under Performance Improvement
Requests 2005-2507 and 2004-0688, and Condition Report 2008-0467 and
found that Wolf Creek designated prior failures nonsignificant.
The failure to take timely corrective actions within 9 years was a performance
deficiency. The inspectors determined that this finding was more than minor
because it is associated with the equipment performance attribute for the
mitigating systems cornerstone; and, it affected the cornerstone objective to
ensure the availability, reliability, and capability of systems that respond to
initiating events to prevent undesirable consequences (i.e. core damage). The
finding screened to Phase 2 because the finding represents an actual loss of
safety function of a single train of high head injection for greater than its
Technical Specification 3.8.1.B.2 allowed outage time of 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. Using an
exposure time of less than 3 days for the scenario Centrifugal Charging Pump
PBG05A [Fails to Run], a bounding risk of Green results from the Phase 2
presolved worksheets. Additionally, the cause of the finding has a human
performance crosscutting aspect in the area associated with resources.
Specifically, Wolf Creek did not ensure adequate resources to maintain long-term
plant safety by minimizing the room coolers long-standing issues and preventive
maintenance deferrals H.2(a) (Section 4OA3.2(iii)).
- Green. The inspectors identified a noncited violation of Technical Specification 5.4.1.d because Procedure OFN RP 017, "Control Room Evacuation," Revision
21, failed to account for the needed actions to reestablish reactor coolant pump
seal cooling. Failure to reestablish seal cooling in a timely manner could have
resulted in a small break loss of coolant accident.
This performance deficiency resulted from an inadequate postfire safe shutdown
procedure. The inspectors determined the finding is greater than minor in that it
affected the ability to achieve and maintain hot shutdown following a control room
fire. This finding is associated with the mitigating systems cornerstone attribute
of protection against external factors (e.g. fire). This finding affected the
mitigating systems cornerstone objective to ensure the availability, reliability, and
capability of systems that respond to external events (such as fire) to prevent
undesirable consequences. In addition to the control room fire requiring
-5- Enclosure
operators to evacuate the control room, the fire would have had to affect
components located in two physically separated panels. The licensee has
Institute of Electrical and Electronics Engineers Standard 383 qualified cables
and conductors throughout the plant. The Phase 3 risk evaluation performed by
the NRC senior reactor analyst determined this deficiency had very low risk
significance (Section 4OA3.5).
- Green. The inspectors identified a noncited violation of License Condition 2.c(5)
because the licensee failed to evaluate the impact of a motor-operated valve
failure mechanism on their ability to implement postfire safe shutdown following a
control room evacuation. The licensee determined that the failure mechanism
affected 38 motor-operated valves and upon valve failure could affect their ability
to implement their postfire safe shutdown procedure. A short circuit that
bypassed the torque and/or limit switches could damage the valves and prevent
repositioning of the valve in the postfire safe shutdown position.
The inspectors determined this was a performance deficiency because the
licensee failed to ensure that components necessary to safely shutdown the
reactor would remain operable following a fire. This deficiency was more than
minor, in that, it had the potential to impact the mitigating systems cornerstone
objective to ensure the availability, reliability, and capability of systems that
respond to external events (such as fire) to prevent undesirable consequences.
In addition to the control room fire requiring operators to evacuate the control
room, the fire would have had to affect components located in five different
control panels. The Phase 3 risk evaluation performed by the NRC senior reactor
analyst determined this deficiency had very low risk significance
(Section 4OA5.2).
Cornerstone: Occupational Radiation Safety
- Green. The inspectors reviewed a self-revealing noncited violation of Technical
Specification 5.7.2.a for failure to evaluate changing radiological conditions and
control an area as a locked high radiation area. Specifically, on October 17,
2007, dose rates in Room 7604 increased to levels requiring posting as a
Locked High Radiation Area, as a result of a vent and drain evolution. Dose
rates reached a level of 1500 mRem/hour prior to the area being properly posted
and controlled. This issue was entered into the licensees corrective action
program as Condition Report 2007-003934. Immediate corrective actions
included posting and controlling the area as a locked high radiation area. Other
corrective actions included changing the vent and drain process to change the
vent path.
This finding is greater than minor because it is associated with the occupational
radiation safety program and process attribute and affected the cornerstone
objective, in that, the failure to properly post and control access to a locked high
radiation area has the potential to increase personnel dose. This occurrence
involves the potential for unplanned, unintended dose. Utilizing Inspection
Manual Chapter 0609, Appendix C, Occupational Radiation Safety Significance
Determination Process, the inspectors determined that the finding was of very
low safety significance because it did not involve; (1) as low as is reasonably
achievable planning and controls, (2) an overexposure, (3) a substantial potential
-6- Enclosure
for overexposure, or (4) an impaired ability to assess dose. This finding has a
crosscutting aspect in the area of human performance associated with the work
control component because licensee failed to appropriately plan work activities
by incorporating job site conditions that may impact radiological safety H.3(a)
(Section 2OS1(i)).
- Green. The inspectors reviewed a self-revealing noncited violation of Technical
Specification 5.4.1 for failure to follow a licensee procedure. Specifically, on
March 29, 2008, one of two radiographers conducting radiography operations in
the quality control vault received a dose rate alarm on their electronic dosimeter.
The two radiographers evaluated the dose received and decided to continue with
radiography without notifying health physics personnel to evaluate the conditions.
This issue was entered into the licensees corrective action program as Condition
Report 2008-001181. Immediate corrective actions included restriction of the
radiographers to log onto the radiation work permit and discussions with the
radiographers and the contractors radiation safety officer. Long-term corrective
action is still being evaluated.
This finding is greater than minor because it is associated with the occupational
radiation safety program and process attribute and affected the cornerstone
objective, in that, the failure to stop work and notify health physics personnel for
assistance had the potential to increase personnel dose. This occurrence
involves the potential for unplanned, unintended dose. Utilizing Inspection
Manual Chapter 0609, Appendix C, Occupational Radiation Safety Significance
Determination Process, the inspector determined that the finding was of very low
safety significance because it did not involve: (1) as low as is reasonably
achievable planning and controls, (2) an overexposure, (3) a substantial potential
for overexposure, or (4) an impaired ability to assess dose. This finding has a
crosscutting aspect in the area of human performance associated with the
decision making component because the radiographer and assistant failed to
contact health physics personnel to discuss the circumstances surrounding the
unexpected dose rate alarm H.1(a) (Section 2OS1(ii)).
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 licensee's corrective action program. These violations and
their corrective actions are listed in Section 4OA7 of this report.
-7- Enclosure
REPORT DETAILS
Summary of Plant Status
The plant started the inspection period at 100 percent rated thermal power. On January 11,
2008, the licensee performed a reactor shutdown due to voiding in emergency core cooling
system (ECCS) piping. After determining the cause and restoring systems to operable, the
plant was returned to full power on January 16, 2008. On March 17, 2008, a 13.8 kV
transformer failure resulted in a plant trip. The plant remained shut down the rest of the report
period and entered Refueling Outage 16 on March 22, 2008.
1. REACTOR SAFETY
Cornerstones: Initiating Events, Mitigating Systems, Barrier Integrity, Emergency
Preparedness
1R01 Adverse Weather Protection (71111.01)
.1 Readiness for Seasonal Susceptibilities
a. Inspection Scope
The inspectors completed a review of the licensee's readiness of seasonal
susceptibilities involving extreme high temperatures. The inspectors: (1) reviewed plant
procedures, the Updated Safety Analysis Report (USAR), and Technical Specifications
to ensure that operator actions defined in adverse weather procedures maintained the
readiness of essential systems; (2) walked down portions of the systems listed below to
ensure that adverse weather protection features were sufficient to support operability
including the ability to perform safe shutdown functions; (3) evaluated operator staffing
levels to ensure the licensee would maintain the readiness of essential systems required
by plant procedures; and (4) reviewed the corrective action program to determine if the
licensee identified and corrected problems related to adverse weather conditions.
- January 23, 2008, cold weather impact on essential service water (ESW)
Documents reviewed are listed in the attachment.
The inspectors completed one sample.
b. Findings
No findings of significance were identified.
.2 Readiness for Impending Adverse Weather Conditions
a. Inspection Scope
The inspectors completed a review of the licensee's readiness for impending adverse
weather involving severe thunderstorms and heavy rains. The inspectors: (1) reviewed
plant procedures, the Updated Safety Analysis Report (USAR), and Technical
Specifications to ensure that operator actions defined in adverse weather procedures
-8- Enclosure
maintained the readiness of essential systems; (2) walked down portions of the systems
listed below to ensure that adverse weather protection features were sufficient to support
operability, including the ability to perform safe shutdown functions; (3) reviewed
maintenance records to determine that applicable surveillance requirements were
current before the anticipated weather developed; and (4) reviewed plant modifications,
procedure revisions, and operator work arounds to determine if recent facility changes
challenged plant operation.
- January 7, 2008, severe thunderstorms caused the loss of two alert notification
system sirens
Documents reviewed are listed in the attachment.
The inspectors completed one sample.
b. Findings
No findings of significance were identified.
1R04 Equipment Alignment (71111.04)
.1 Quarterly Partial System Walkdowns
a. Inspection Scope
The inspectors performed partial system walkdowns of the following risk-significant
systems:
- February 11, 2008, Sharpe station alignment to Wolf Creek switchyard
- March 6, 2008, emergency diesel Generator (EDG) B while ESW A is inoperable
- March 12, 2008, motor-driven auxiliary feedwater Pump A while turbine-driven
auxiliary feedwater (TDAFW) is inoperable
The inspectors selected these systems based on their risk significance relative to the
reactor safety cornerstones at the time they were inspected. The inspectors attempted
to identify any discrepancies that could impact 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 (WOs), 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 walked down accessible portions of the systems to verify system
components and support equipment were aligned correctly and operable. The
inspectors examined the material condition of the components and observed operating
parameters of equipment to verify that there were no obvious deficiencies. The
inspectors also verified that the licensee had properly identified and resolved equipment
alignment problems that could cause initiating events or impact the capability of
mitigating systems or barriers and entered them into the corrective action program with
the appropriate significance characterization.
-9- Enclosure
Documents reviewed are listed in the attachment.
The inspectors completed three samples.
b. Findings
No findings of significance were identified.
1R05 Fire Protection (71111.05)
.1 Routine Resident Inspector Tours (71111.05Q)
a. Inspection Scope
The inspectors conducted fire protection walkdowns which were focused on availability,
accessibility, and the condition of firefighting equipment in the following risk-significant
plant areas:
- January 17, 2008, auxiliary building 1988 elevation
- January 31, 2008, auxiliary building 2026 elevation
- February 7, 2008, control building 2000 elevation
- March 10, 2007, turbine building 2037 elevation
The inspectors reviewed areas to assess if the licensee 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 (IPEEE)
with later additional insights, their potential to impact equipment which could initiate or
mitigate a plant transient, or their impact on the plants ability to respond to a security
event. Using the documents listed in the attachment, the inspectors verified that fire
hoses and extinguishers were in their designated locations and available for immediate
use; that fire detectors and sprinklers were unobstructed, that transient material loading
was within the analyzed limits; and fire doors, dampers, and penetration seals appeared
to be in satisfactory condition. The inspectors also verified that minor issues identified
during the inspection were entered into the licensees corrective action program.
Documents reviewed by the inspectors are listed in the attachment.
The inspectors completed four samples
b. Findings
Introduction. The inspectors identified a Green noncited violation (NCV) of Technical
Specification 5.4.1.d for failure to implement fire protection impairment control permit
- 10 - Enclosure
requirements and compensatory measures following a trouble alarm for a fire detector in
the auxiliary building.
Description. On January 25, 2008, a fire protection trouble alarm came in on the fire
alarm control Panel KC-008. The control room supervisor acknowledged the alarm
associated with Detector KC-104-XCH-ID-006. Alarm Procedure ALR KC-008, Fire
Protection Panel KC-008 Alarm Response, Revision 15, Step 4.3.1 requires, in part,
that a fire protection impairment permit be initiated for a smoke detector trouble alarm.
The control room supervisor reviewed the impairment log and found
Impairment 2008-020 for a detector in the same location as the alarm. Based on this
information, the control room supervisor did not initiate an impairment request. Alarming
Detector KC-104-XCH-ID-006 was adjacent to Detector KC-104-XSH-ID-007 that was
listed in Impairment 2008-020. However, the control room supervisor did not verify the
alarming detector point was the same detector point listed in the impairment.
On January 26, 2008, while performing Procedure STN KC-008, Fire Alarm Control
Panel KC-008 Daily Check, Revision 7, which required operators to check KC-008
alarms and trouble points, it was discovered that Detector KC-104-XCH-ID-006 did not
have a fire protection impairment control permit. This detector was adjacent to
Detector KC-04-XSH-ID-07 which was already inoperable in Impairment 2008-020.
Administrative Procedure AP 10-103, Fire Protection Impairment Control, Revision 22,
required fire detection in the area to be declared inoperable if two adjacent detectors are
inoperable. This condition existed for approximately 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and required a
compensatory continuous fire watch for the period that both detectors were inoperable.
Upon discovery, a continuous fire watch was stationed and emergent work was declared
to clean the dirty detectors.
Procedure AP 21-001, Conduct of Operations, Revision 39, requires, in part, that watch
stations are responsible for reviewing and understanding pertinent information such as
control room turnover checklists for shift relief and turnover. The procedure also states
that shift turnover discussions should include reasons for alarms and status panel lights.
The residents noted that the control room turnover checklist contains a section for listing
the KC008 alarms; however, the two turnover checklists for the two shifts following the
initial alarm did not identify Detector KC-104-XCH-ID-006 as a KC-008 alarm. The
control room turnover checklist also has specific requirements to review fire protection
permits and verify that KC-008 alarms are not disabled or disarmed without
documentation. However, neither review discovered that the alarm did not have an
impairment issued. Based on this information, the residents concluded that the licensee
had two previous opportunities to identify the condition during control room turnovers.
After reviewing the licensees evaluation of the condition, the residents noted that these
aspects were not identified in the evaluations conclusions or corrective actions which
focused on only the initial error performed by the operator. During interviews with
control room operators, the inspectors noted that operators are trained to ask for and get
peer checks for verification of alarms and disabled points but failed to utilize any human
error prevention tools in this instance.
Analysis. The failure to implement fire protection impairment control permit
requirements and establish compensatory measures for the auxiliary building 2026' level
was considered a performance deficiency. Traditional enforcement does not apply since
there were no actual safety consequences or potential for impacting the NRCs
- 11 - Enclosure
regulatory function, and the finding was not the result of any willful violation of NRC
requirements or Wolf Creek procedures. The inspectors determined that the finding was
more than minor because it was associated with the mitigating systems cornerstone
attribute of protection against external factors and affected the cornerstone objective of
ensuring the availability of systems that respond to initiating events to prevent
undesirable consequences. Specifically, this issue relates to the protection against fire
example of protection against external factors attribute because the detectors were
inoperable without ensuring compensatory measures where in place. The inspectors
evaluated the significance of this finding using Phase 1 of Inspection Manual
Chapter (IMC) 0609, Appendix F, "Fire Protection Significance Determination Process,
the inspectors determined that the finding was of very low safety significance because it
involved compensatory measures for the fixed fire protection system and was assigned
a low degradation rating since less than 10 percent of the fire detectors in the area were
disabled. The inspectors also determined that the finding has crosscutting aspects in
the area of human performance associated with work practices because the licensee
failed to apply appropriate human error techniques such as self- and peer-checking
techniques to avoid committing errors H.4(a).
Enforcement. Technical Specification 5.4.1.d requires that written procedures be
established, implemented, and maintained covering activities related to fire protection
program implementation. Administrative Procedure AP 10-103, Fire Protection
Impairment Control, Revision 21, requires, in part, fire protection impairment control
permit shall be prepared in order to determine the appropriate compensatory measures
and track the impairment. Contrary to the above, on January 25, 2008, two fire
detectors were inoperable in the auxiliary building 2026' level without implementing a fire
protection impairment control permit and establishing compensatory measures. This
issue and the corrective actions are being tracked by the licensee in Condition
Report (CR) 2008-001657. Because the finding is of very low safety significance and
has been entered into the corrective action program, this violation is being treated as an
NCV 05000482/2008002-01, Failure to Implement Fire Protection Impairment Control
Permit Requirements and Compensatory Measures.
1R11 Licensed Operator Requalification Program (71111.11)
Resident Inspector Quarterly Review (71111.11Q)
a. Inspection Scope
The inspectors observed testing and training of senior reactor operators and reactor
operators to identify deficiencies and discrepancies in the training, to assess operator
performance, and to assess the evaluator's critique. The training scenario involved:
- February 1, 2008, loss of residual heat removal (RHR) during shutdown
conditions
Documents reviewed by the inspectors are listed in the attachment.
The inspectors completed one sample.
- 12 - Enclosure
b. Findings
No findings of significance were identified.
1R12 Maintenance Effectiveness (71111.12)
.1 Routine Quarterly Evaluations (71111.12Q)
a. Inspection Scope
The inspectors reviewed the maintenance activities listed below to: (1) verify the
appropriate handling of structure, system, and component (SSC) performance or
condition problems; (2) verify the appropriate handling of degraded SSC functional
performance; (3) evaluate the role of work practices and common cause problems; and
(4) evaluate the handling of SSC issues reviewed under the requirements of the
maintenance rule, 10 CFR Part 50, Appendix B, and Technical Specifications.
- November 27, 2007, service water Pump A trip due to SL41 bus transients
Documents reviewed by the inspectors are listed in the attachment.
The inspectors completed one sample.
b. Findings
No findings of significance were identified.
1R13 Maintenance Risk Assessments and Emergent Work Control (71111.13)
a. Inspection Scope
Risk Assessment and Management of Risk
The inspectors reviewed the licensee'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:
- March 17-22, 2008, weekly T-0 risk assessment profile
- February 12 and 13, 2008, STS IC-805B degraded grid voltage relay testing
These activities were selected based on their potential risk significance relative to the
reactor safety cornerstones. As applicable for each activity, the inspectors verified that
risk assessments were performed as required by 10 CFR 50.65(a)(4) and were accurate
and complete. When emergent work was performed, the inspectors verified that the
plant risk was promptly reassessed and managed. The inspectors reviewed the scope
of maintenance work, discussed the results of the assessment with the licensee's
probabilistic risk analyst or shift technical advisor, and verified plant conditions were
consistent with the risk assessment. The inspectors also reviewed Technical
Specification requirements and walked down portions of redundant safety systems,
- 13 - Enclosure
when applicable, to verify risk analysis assumptions were valid and applicable
requirements were met.
Emergent Work Control
For the emergent work activities listed below, the inspectors: (1) verified that the
licensee performed actions to minimize the probability of initiating events and maintained
the functional capability of mitigating systems and barrier integrity systems; (2) verified
that emergent work-related activities such as troubleshooting, work planning/scheduling,
establishing plant conditions, aligning equipment, tagging, temporary modifications, and
equipment restoration did not place the plant in an unacceptable configuration; and
(3) reviewed the corrective action program to determine if the licensee identified and
corrected risk assessment and emergent work control problems.
- January 11, 2008, shutdown due to ECCS voiding
- March 11, 2008, scaffolding installation resulting in reactive load swings
Documents reviewed by the inspectors are listed in the attachment.
The inspectors completed four samples.
b. Findings
Introduction. The inspectors identified a Green NCV of Technical Specification 3.8.1.B.4
in which the licensee removed equipment from service that was prohibited by Technical
Specifications.
Description. On February 11, 2008, the licensee entered TS 3.8.1.B.4.2.2. This
specification allowed an emergency diesel generator to be inoperable for up to 7 days
during an once per cycle extended outage. On February 12 and 13, inspectors
observed surveillance Procedure STS IC-805B, Channel Calibration of NB02 Grid
Degraded Voltage, Time Delay Trip, Revision 11, for testing of the Train B of degraded
voltage relays for offsite power. Offsite power Train B was declared inoperable at
10:09 a.m. on February 12. On February 12, instrumentation and control (I&C)
technicians partially completed STS 805B, but required clarification of their procedure
and secured the test and restored the equipment to operable status. On February 13,
offsite power Train B was again declared inoperable at 7:49 a.m. Inspectors reviewed
Technical Specification Bases 3.8.1.B.4 which prohibits elective maintenance within the
switchyard that would challenge offsite power while in the 7-day emergency diesel
generator extended outage. Inspectors also reviewed the NRC Safety Evaluation
Report (SER) for the 7 day EDG allowed outage time (Technical
Specification 3.8.1.B.4.2.2) and found that Section 4.6.c, states: The offsite power
supply and switchyard conditions are conducive to an extend[ed] DG [completion time],
which includes ensuring that switchyard access is restricted and no elective
maintenance within the switchyard is performed that would challenge the offsite power
availability. Additionally, Condition D of the Technical Specification Bases states that
no equipment or systems assumed to be available for the extended EDG completion
time are removed from service, which includes auxiliary feedwater, component cooling
water, essential service water and their support systems. However, Wolf Creek
removed one train of offsite power degraded voltage relays which affects offsite power to
- 14 - Enclosure
Bus NB02 (Train B) which is a support system for the above equipment. The inspectors
found that Procedure STS IC 805B permits the testing of degraded voltage relays only
while the diesel is out of service. The inspectors determined that this practice is
acceptable when performing offsite power maintenance under Technical Specification 3.8.1.B.4.1, but not Technical Specification 3.8.1.B.4.2.2 due to the
increase in risk for the longer allowed outage period. Procedure STS IC-805B was not
revised subsequent to issuance of License Amendment 163 and permitted the work to
occur. Additionally, Procedure AP 22C-003, Operational Risk Assessment Program,
Revision 13, prohibits elective maintenance within the switchyard that would challenge
offsite power during Technical Specification 3.8.1.B.4.2.2. Wolf Creek appropriately
restricted access to the portion of the switchyard outside the protected area but did not
appropriately restrict work for offsite power inside the protected area. The inspectors
determined that challenges to offsite power can originate with elective maintenance
inside the protected area. Inspectors found that Wolf Creek assessed risk under
10 CFR 50.65 a(4) for this evolution which resulted in elevating risk to yellow during
testing.
Analysis. The inspectors determined that the failure to follow the NRC SER and
Technical Specification Bases for Technical Specification 3.8.1.B.4 was a performance
deficiency. Traditional enforcement does not apply since there were no actual safety
consequences or potential for impacting the NRC's regulatory function, and the finding
was not the result of any willful violation of NRC requirements or Wolf Creek procedures.
The inspectors determined that this finding was more than minor because it is
associated with the equipment performance attribute for the mitigating systems
cornerstone; and, it affected the cornerstone objective to ensure the availability,
reliability, and capability of systems that respond to initiating events to prevent
undesirable consequences (i.e. core damage). Specifically, this issue relates to the
availability and reliability examples of the equipment performance attribute because an
offsite power source was at greater risk of being lost.
The inspectors evaluated the significance of this finding using Phase 1 of IMC 0609,
Appendix A, Significance Determination of Reactor Inspection Findings for At-Power
Situations, and determined that the finding was of very low safety significance because
the issue resulted in the Train B offsite power being inoperable, but capable of supplying
the safety bus for greater than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. As such, under Phase 1 screening, the
deficiency is not related to a qualification or design deficiency, it did not represent a loss
of safety function for a train or system as defined in the plant specific risk-informed
inspection notebook, and was not related to external events such as fires and floods.
Additionally, the cause of the finding has human performance crosscutting aspects in the
area associated with work control. Specifically, Wolf Creek did not ensure STS IC-805B
was appropriately coordinated within organizations to assure plant and human
performance during the extended EDG allowed outage time H.3(b).
Enforcement. Technical Specification 3.8.1, Condition B.4.2.2, permits one diesel
generator to be inoperable for 7 days provided the limitations articulated in the NRC
SER for License Amendment 163. The NRC SER requires that the offsite power supply
and switchyard conditions be conducive to an extend[ed] diesel generator [completion
time], which includes ensuring that switchyard access is restricted and no elective
maintenance within the switchyard is performed that would challenge the offsite power
availability. The NRC SER also requires support equipment to systems such as auxiliary
feedwater, component cooling water, and ESW to be available. Contrary to the above,
- 15 - Enclosure
on February 12 and 13, 2008, Wolf Creek performed elective maintenance on the
Train B offsite power degraded voltage relays while the Train B emergency diesel
generator was in an extended outage. Because the finding is of very low safety
significance and has been entered into the corrective action program as CR 2008-
001675, this violation is being treated as an NCV, consistent with Section VI.A of the
NRC Enforcement Policy: NCV 05000482/2008002-02, Performing Prohibited Elective
Maintenance on Offsite Power During EDG Maintenance.
1R15 Operability Evaluations (71111.15)
a. Inspection Scope
The inspectors: (1) reviewed plants status documents such as operator shift logs,
emergent work documentation, deferred modifications, and standing orders to determine
if an operability evaluation was warranted for degraded components; (2) referred to the
USAR and design basis documents to review the technical adequacy of licensee
operability evaluations; (3) evaluated compensatory measures associated with
operability evaluations; (4) determined degraded component impact on any Technical
Specifications; (5) used the significance determination process to evaluate the risk
significance of degraded or inoperable equipment; and (6) verified that the licensee has
identified and implemented appropriate corrective actions associated with degraded
components.
- January 22, 2008, containment sump fabrication and calculation errors
- February 13, 2008, CCP A room cooler leak
- February 28, 2008, ECCS voids
- March 11, 2008, safety injection tank nitrogen leak
Documents reviewed by the inspectors are listed in the attachment.
The inspectors completed four samples.
b. Findings
An unresolved item (URI) was identified when an operability determination dated
January 22, 2008, was required to ensure that latent fabrication and calculation errors
did not create unacceptable reductions in net positive suction head requirements for
pumps in emergency core cooling systems. This new design information was
associated with the already installed containment recirculation sump strainer
modification. The associated vendor calculation, TD 6002 05, for clean strainer head
loss omitted the head loss component associated with the orifices located in the strainer
support plate. The size of the orifice beneath each strainer tube was not large enough to
prevent head loss in excess of the net positive suction head required per the design
conditions defined in the purchase specification supplied to the strainer vendor. The
additional head loss due to the calculation correction was 2.28 feet. This resulted in
required net positive suction head being less than available. Wolf Creek performed an
operability determination review to demonstrate that the head loss margin could be
recovered. The operability determination on January 22, 2008, addressed the smaller
- 16 - Enclosure
support plate orifice holes by using additional margin gained by separating the head loss
of the RHR and containment spray piping systems to demonstrate lower losses and
additional water inventory in containment prior to containment spray swapover to the
sump. Wolf Creek is replacing the strainer support plate with larger orifices to regain
head loss margin in Refueling Outage 16. However, additional concerns were provided
to the licensee by the vendor on April 1, 2008, addressing nonconservative temperature
correction through the orifices. Subsequent to this, the licensee will need to perform
additional analyses to determine if negative margin existed during the last cycle and if
the new orifice holes will provide positive margin. At the completion of the inspection
period, there were still unresolved questions about the assumptions and results
associated with the calculations used for regaining net positive suction head margin.
These concerns require additional inspection and, when completed, the inspection
results will require significance determination. This issue is considered unresolved
pending additional NRC review of Wolf Creek operability determination calculations:
URI 05000483/2008002-03, Containment Sump Net Positive Suction Head Losses.
1R18 Plant Modifications (71111.18)
.1 Permanent Modification Review
a. The inspectors reviewed key affected parameters associated with energy needs,
materials/replacement components, timing, heat removal, control signals, equipment
protection from hazards, operations, flowpaths, pressure boundary, ventilation boundary,
structural, process medium properties, licensing basis, and failure modes for the one
modification listed below. The inspectors verified that: (1) modification preparation,
staging, and implementation does not impair emergency/abnormal operating procedure
actions, key safety functions, or operator response to loss of key safety functions;
(2) postmodification testing will maintain the plant in a safe configuration during testing
by verifying that unintended system interactions will not occur, SSC performance
characteristics still meet the design basis, the appropriateness of modification design
assumptions, and the modification test acceptance criteria has been met; and (3) the
licensee has identified and implemented appropriate corrective actions associated with
permanent plant modifications
- March 6, 2008, containment spray recirculation piping for full flow testing
Documents reviewed by the inspectors are listed in the attachment.
The inspectors completed one sample.
b. Findings
No findings of significance were identified.
.2 Temporary Modification Review
a. Inspection Scope
The inspectors reviewed plant drawings, procedure requirements, and Technical
Specifications to ensure that the below temporary modification was properly
implemented. The inspectors: (1) verified that the modification did not have an affect on
- 17 - Enclosure
system operability/availability, (2) verified that the installation was consistent with the
modification documents, (3) ensured that the post installation test results were
satisfactory and that the impact of the temporary modification on permanently installed
SSCs were supported by the test, (4) verified that the modifications were identified on
control room drawings and that appropriate identification tags were placed on the
affected drawings, and (5) verified that appropriate safety evaluations were completed.
The inspectors verified that licensee identified and implemented any needed corrective
actions associated with temporary modifications.
- January 22, 2008, safety injection room cooler temporary modification procedure
- February 6, 2008, rod control circuitry monitoring equipment for troubleshooting
Documents reviewed by the inspectors are listed in the attachment.
The inspectors completed two samples.
b. Findings
No findings of significance were identified
1R19 Postmaintenance Testing (71111.19)
a. Inspection Scope
The inspectors selected the below listed postmaintenance test activities of risk
significant systems or components. For each item, the inspectors: (1) reviewed the
applicable licensing basis and/or design-basis documents to determine the safety
functions; (2) evaluated the safety functions that may have been affected by the
maintenance activity; and (3) reviewed the test procedure to ensure it adequately tested
the safety function that may have been affected. The inspectors either witnessed or
reviewed test data to verify that acceptance criteria were met, plant impacts were
evaluated, test equipment was calibrated, procedures were followed, jumpers were
properly controlled, the test data results were complete and accurate, the test equipment
was removed, the system was properly realigned, and deficiencies during testing were
documented. The inspectors also reviewed the USAR and corrective action program to
determine if the licensee identified and corrected problems related to postmaintenance
testing.
- January 31, 2008, safety injection Pump A run following planned maintenance
- February 15, 2008, EDG B run following planned maintenance
- March 5, 2008, centrifugal charging Pump A following planned maintenance
Documents reviewed by the inspectors are listed in the attachment.
The inspectors completed three samples.
- 18 - Enclosure
b. Findings
No findings of significance were identified.
1R20 Outage Activities (71111.20)
.1 Refueling Outage Activities
a. Inspection Scope
The inspectors reviewed the outage safety plan and contingency plans for Wolf Creek
Refueling Outage 16 that started on March 22, 2008, and continued into the next period,
to confirm that the licensee had appropriately considered risk, industry experience, and
previous site-specific problems in developing and implementing a plan that assured
maintenance of defense indepth. During the refueling outage, the inspectors observed
portions of the shutdown and cooldown processes and monitored licensee controls over
the outage activities listed below.
- Licensee configuration management, including maintenance of defense indepth
commensurate with the outage safety plan for key safety functions and
compliance with the applicable Technical Specifications when taking equipment
out of service.
- Implementation of clearance activities and confirmation that tags were properly
hung and equipment appropriately configured to safely support the work or
testing.
- Installation and configuration of reactor coolant pressure, level, and temperature
instruments to provide accurate indication, accounting for instrument error.
- Controls over the status and configuration of electrical systems to ensure that
Technical Specification and outage safety plan requirements were met, and
controls over switchyard activities.
- Monitoring of decay heat removal processes, systems, and components.
- Controls to ensure 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 Technical Specifications.
- Refueling activities, including fuel handling and sipping to detect fuel assembly
leakage.
- 19 - Enclosure
- Licensee identification and resolution of problems related to refueling outage
activities.
The inspectors will complete this sample in the next quarter.
b. Findings
No findings of significance were identified.
.2 Other Outage Activities
a. Inspection Scope
The inspectors evaluated forced outage activities for an unscheduled outage that began
on January 11, 2008, and continued through January 16, 2008, due to a Technical
Specification required shutdown for voiding in ECCSs. The inspectors reviewed
activities to ensure that the licensee considered risk in developing, planning, and
implementing the outage schedule. The inspectors observed or reviewed the reactor
shutdown and cooldown, outage equipment configuration, risk management, electrical
lineups, selected clearances, control and monitoring of decay heat removal, control of
containment activities, and identification and resolution of problems associated with the
outage. The inspectors observed portions of the reactor startup and heatup.
Documents reviewed by the inspectors are listed in the attachment.
The inspectors completed one sample.
b. Findings
No findings of significance were identified for the forced outage. Aspects of the ECCS
voiding will be contained in NRC Special Inspection Report 05000482/2008-008.
.3 Other Outage Activities
a. Inspection Scope
The inspectors evaluated forced outage activities for an unscheduled outage that began
on March 17, 2008, due to a reactor trip due to XPB03 transformer failure, and continued
through the start of Refueling Outage 16. The inspectors reviewed activities to ensure
that the licensee considered risk in developing, planning, and implementing the outage
schedule. The inspectors observed or reviewed the reactor shut down and cool down,
outage equipment configuration and risk management, electrical lineups, selected
clearances, control and monitoring of decay heat removal, control of containment
activities, and identification and resolution of problems associated with the outage.
Documents reviewed by the inspectors are listed in the attachment.
The inspectors completed one sample.
- 20 - Enclosure
b. Findings
No findings of significance were identified for the forced outage. Additional aspects of
the plant trip are in Section 4OA3.3.
1R22 Surveillance Testing (71111.22)
.1 Routine Surveillance Testing
a. Inspection Scope
The inspectors reviewed the test results for the following activities to determine whether
risk-significant systems and equipment were capable of performing their intended safety
function and to verify testing was conducted in accordance with applicable procedural
and Technical Specification requirements:
- January 26, 2008, emergency exhaust system run
- February 11-15, 2008, offsite power alignment Technical Specification
surveillance
- March 6, 2008, EDG A biennial 24-hour endurance and load test
The inspectors observed in-plant activities and reviewed procedures and associated
records to determine whether: any preconditioning occurred; effects of the testing were
adequately addressed by control room personnel or engineers prior to the
commencement of the testing; acceptance criteria were clearly stated, demonstrated
operational readiness, and were consistent with the system design basis; plant
equipment calibration was correct, accurate, and properly documented; as left setpoints
were within required ranges; the calibration frequency was in accordance with Technical
Specifications, the USAR, procedures, and applicable commitments; measuring and test
equipment calibration was current; test equipment was used within the required range
and accuracy; applicable prerequisites described in the test procedures were satisfied;
test frequencies met Technical Specification requirements to demonstrate operability
and reliability; tests were performed in accordance with the test procedures and other
applicable procedures; jumpers and lifted leads were controlled and restored where
used; test data and results were accurate, complete, within limits, and valid; test
equipment was removed after testing; where applicable, test results not meeting
acceptance criteria were addressed with an adequate operability evaluation or the
system or component was declared inoperable; where applicable for safety-related
instrument control surveillance tests, reference setting data were accurately incorporated
in the test procedure; where applicable, actual conditions encountering high resistance
electrical contacts were such that the intended safety function could still be
accomplished; prior procedure changes had not provided an opportunity to identify
problems encountered during the performance of the surveillance or calibration test;
equipment was returned to a position or status required to support the performance of
the safety functions; and all problems identified during the testing were appropriately
documented and dispositioned in the corrective action program.
- 21 - Enclosure
Documents reviewed by the inspectors are listed in the attachment.
The inspectors completed four samples
b. Findings
No findings of significance were identified.
.2 In-service Testing Surveillance
a. Inspection Scope
The inspectors reviewed the test results for the following activities to determine whether
risk-significant systems and equipment were capable of performing their intended safety
function and to verify testing was conducted in accordance with applicable procedural
and Technical Specification requirements:
- January 23, 2008, inservice testing of residual heat removal Pump B
- February 5, 2008, inservice testing of atmospheric relief Valve D
The inspectors observed in-plant activities and reviewed procedures and associated
records to determine whether: any preconditioning occurred; effects of the testing were
adequately addressed by control room personnel or engineers prior to the
commencement of the testing; acceptance criteria were clearly stated, demonstrated
operational readiness, and were consistent with the system design basis; plant
equipment calibration was correct, accurate, and properly documented; as left setpoints
were within required ranges; and the calibration frequency were in accordance with
Technical Specifications, the USAR, procedures, and applicable commitments;
measuring and test equipment calibration was current; test equipment was used within
the required range and accuracy; applicable prerequisites described in the test
procedures were satisfied; test frequencies met Technical Specification requirements to
demonstrate operability and reliability; tests were performed in accordance with the test
procedures and other applicable procedures; jumpers and lifted leads were controlled
and restored where used; test data and results were accurate, complete, within limits,
and valid; test equipment was removed after testing; where applicable for inservice
testing activities, testing was performed in accordance with the applicable version of
Section XI, American Society of Mechanical Engineers (ASME) Code, and reference
values were consistent with the system design basis; where applicable, test results not
meeting acceptance criteria were addressed with an adequate operability evaluation or
the system or component was declared inoperable; where applicable for safety-related
instrument control surveillance tests, reference setting data were accurately incorporated
in the test procedure; where applicable, actual conditions encountering high resistance
electrical contacts were such that the intended safety function could still be
accomplished; prior procedure changes had not provided an opportunity to identify
problems encountered during the performance of the surveillance or calibration test;
equipment was returned to a position or status required to support the performance of its
safety functions; and all problems identified during the testing were appropriately
documented and dispositioned in the corrective action program. Documents reviewed
are listed in the attachment.
- 22 - Enclosure
Documents reviewed by the inspectors are listed in the attachment.
The inspectors completed two samples.
b. Findings
No findings of significance were identified.
1EP6 Drill Evaluation (71114.06)
a. Inspection Scope
The drill listed below contributed to drill/exercise performance and emergency response
organization performance indicators. The inspectors: (1) observed the training evolution
to identify any weaknesses and deficiencies in classification, notification, and protective
action requirements development activities; (2) compared the identified weaknesses and
deficiencies against licensee identified findings to determine whether the licensee is
properly identifying failures; and (3) determined whether licensee performance is in
accordance with the guidance of the Nuclear Energy Institute (NEI) 99-02 documents
acceptance criteria.
- January 31, 2008, loss of all annunciators followed by loss of all offsite power
Documents reviewed by the inspectors are listed in the attachment.
The inspectors completed one sample.
b. Findings
No findings of significance were identified.
2. RADIATION SAFETY
Cornerstone: Occupational Radiation Safety
2OS1 Access Control to Radiologically Significant Areas (71121.01)
Review of Licensee Performance Indicators for the Occupational Exposure Cornerstone
a. Inspection Scope
This area was inspected to assess the licensees performance in implementing physical
and administrative controls for airborne radioactivity areas, radiation areas, high
radiation areas, and worker adherence to these controls. The inspectors used the
requirements in 10 CFR Part 20, the Technical Specifications, and the licensees
procedures required by Technical Specifications as criteria for determining compliance.
During the inspection, the inspectors interviewed the radiation protection manager,
radiation protection supervisors, and radiation workers. The inspectors performed
independent radiation dose rate measurements and reviewed the following items:
- Performance indicator events and associated documentation packages reported
by the licensee in the occupational radiation safety cornerstone
- 23 - Enclosure
- Controls (surveys, posting, and barricades) of radiation, high radiation, or
airborne radioactivity areas
- Radiation work permits, procedures, engineering controls, and air sampler
locations
- Conformity of electronic personal dosimeter alarm set points with survey
indications and plant policy; workers knowledge of required actions when their
electronic personnel dosimeter noticeably malfunctions or alarms
- Barrier integrity and performance of engineering controls in airborne radioactivity
areas
- Physical and programmatic controls for highly activated or contaminated
materials (nonfuel) stored within spent fuel and other storage pools.
- Self-assessments, audits, licensee event reports (LERs), and special reports
related to the access control program since the last inspection
- Corrective action documents related to access controls
- Licensee actions in cases of repetitive deficiencies or significant individual
deficiencies
- Radiation work permit briefings and worker instructions
- Adequacy of radiological controls, such as required surveys, radiation protection
job coverage, and contamination control during job performance
- Dosimetry placement in high radiation work areas with significant dose rate
gradients
- Changes in licensee procedural controls of high dose rate - high radiation areas
and very high radiation areas
- Controls for special areas that have the potential to become very high radiation
areas during certain plant operations
- Posting and locking of entrances to all accessible high dose rate - high radiation
areas and very high radiation areas
- Radiation worker and radiation protection technician performance with respect to
radiation protection work requirements
Documents reviewed by the inspector are listed in the attachment.
The inspector completed 20 of the required 21 samples.
- 24 - Enclosure
b. Findings
.1 Introduction. A self-revealing NCV of Technical Specification 5.7.2.a was reviewed
involving the failure to evaluate changing radiological conditions in Room 7604 and
control the area as a locked high radiation area. The violation had very low safety
significance.
Description. On September 26, 2007, mixed bed Demineralizer A was sluiced to the
primary spent resin storage tank. During a followup survey, a localized concentration of
resin was identified in the common sluice header in the 2022 pipe chase of the
radioactive waste building. Operations was unable to perform a lineup to remove the
resin because a procedure for that evolution did not exist. It was decided to leave the
resin in place since it was in a remote location, the area was being controlled as a
locked high radiation area, and it would be removed during the next resin sluice
evolution.
On October 15, 2007, Clearance Order EC-N-005 was developed to vent and drain a
section of piping in preparation for work on Valve ECV0081. The vent path for
Clearance Order EC-N-005 was through Valve ECV0079, located in Room 7604, which
ties into the common resin sluice header. Dose rates in Room 7604 are typically in the
range of 8 - 10 mRem/hour. The clearance order group was not informed of the
localized concentration of resin remaining in the sluice header. In preparation for
hanging the clearance, operations radioactive waste personnel and health physics
personnel discussed the dose rates in the affected areas, valve manipulations, and that
the vent path was hard piped and would not require a temporary hose connection.
On the morning of October 16, 2007, a radwaste person performing routine evolutions
entered Room 7604 and immediately received a dose rate alarm on his electronic
dosimeter. The operator immediately exited the room and notified health physics
personnel. An evaluation of the individuals electronic dosimeter indicated he had
entered an area with a dose rate of 74 mRem/hour. The dosimeter had been set to
alarm at 50 mRem/hour. Radiological surveys of the area taken at 9 a.m. on October 16
indicated dose rates as high as 197 mRem/hour. The area was immediately posted as a
high radiation area. At this time, the licensee did not understand the cause of the
increased radiation levels. Followup surveys were taken at 9 a.m. on October 17 and
indicated that dose rates had increased to 1500 mRem/hour requiring posting and
control as a locked high radiation area. The area was immediately posted and controlled
as a locked high radiation area. Subsequent surveys showed dose rates reached a
maximum of 2500 mRem/hour before a temporary instruction was written to flush the
resin from the common sluice header.
The inspectors determined that health physics personnel failed to perform timely surveys
to identify and control a locked high radiation area. Corrective actions included
immediately posting and controlling the area as a locked high radiation area and
developing a temporary procedure to flush the resin from the common sluice header to
the spent resin storage tank.
Analysis. This finding is more minor because it is associated with the occupational
radiation safety program and process attribute and affected the cornerstone objective, in
that, the failure to properly post and control access to a locked high radiation area has
the potential to increase personnel dose. This occurrence involves the potential for
- 25 - Enclosure
unplanned, unintended dose. Utilizing IMC 0609, Appendix C, Occupational Radiation
Safety Significance Determination Process, the inspector determined that the finding
was of very low safety significance because it did not involve: (1) as low as is
reasonably achievable (ALARA) planning and controls, (2) an overexposure, (3) a
substantial potential for overexposure, or (4) an impaired ability to assess dose. This
finding has a crosscutting aspect in the area of human performance associated with the
work control component because the licensee failed to appropriately plan work activities
by incorporating job site conditions that may impact radiological safety H.3(a).
Enforcement. Technical Specification 5.7.2.a requires that the entryway to high radiation
areas with dose rates greater than 1.0 Rem/hour be conspicuously posted as a high
radiation area and shall be provided with a locked or continuously guarded door or gate.
Contrary to this requirement, the licensee failed to perform a timely survey of Room
7604 and evaluate changing radiological conditions which, required the room to be
posted and controlled as a Locked High Radiation Area. Because the finding is of very
low safety significance and has been entered into the licensees corrective action
program as CR 2007-003934, this violation is being treated as an NCV consistent with
Section VI.A of the Enforcement Policy: NCV 05000482/2008002-04, Failure to Control
Area as a Locked High Radiation Area.
.2 Introduction. The inspectors reviewed a self-revealing NCV of Technical
Specification 5.4.1 for failure to follow a licensee procedure.
Description. On March 29, 2008, while performing radiography at the quality control
vault, a radiographer assistant received a dose rate alarm on his electronic dosimeter.
Radiography evolutions at the site are controlled using a radiation work permit provided
by the health physics department. Radiation Work Permit 08-3021 established a dose
rate alarm setpoint of 500 mRem/hour. The radiography crew properly secured the
radiography source and performed surveys of the source camera. The radiographer and
assistant reviewed the dose received by each individual as indicated on their electronic
dosimeters and, without notifying health physics personnel, decided to continue with
radiography. The alarm condition was noted when the radiographer and assistant
returned to access control to sign off of the radiation work permit. A review of the
dosimeters indicated that the assistant received a total dose of 2.0 mRem with a peak
dose rate of 512 mRem/hour and the radiographer received 2.9 mRem with a peak dose
rate of 476 mRem/hour. Immediate corrective actions included restriction of the
radiographers to log onto the radiation work permit and discussions with the
radiographers and the contractors radiation safety officer. Long-term corrective action is
still being evaluated.
Analysis. This finding is greater than minor because it is associated with the
occupational radiation safety program and process attribute and affected the
cornerstone objective, in that the failure to stop work and notify health physics personnel
for assistance had the potential to increase personnel dose. This occurrence involves
the potential for unplanned, unintended dose. Utilizing IMC 0609, Appendix C,
Occupational Radiation Safety Significance Determination Process, the inspectors
determined that the finding was of very low safety significance because it did not involve:
(1) ALARA planning and controls, (2) an overexposure, (3) a substantial potential for
overexposure, or (4) an impaired ability to assess dose. This finding has a crosscutting
aspect in the area of human performance associated with the decision making
component because the radiographer and assistant failed to contact health physics
- 26 - Enclosure
personnel to discuss the circumstances surrounding the unexpected dose rate alarm
Enforcement. Technical Specification 5.4.1 requires procedures be established,
implemented, and maintained covering the applicable procedures recommended in
Regulatory Guide 1.33, Appendix A. Section 7 of Appendix A recommends radiation
protection procedures for personnel monitoring. Licensee Procedure AP 25B-100,
Radiation Worker Guidelines, Section 6.2.8 states, in part, If an individuals electronic
dosimeter alarms, the worker shall notify coworkers/health physics and exit the area.
Health physics personnel will then evaluate radiological conditions prior to the
continuation of work. Contrary to this requirement, the radiographer and assistant failed
to notify health physics personnel prior to resuming work activities. Because this failure
to follow a procedure is of very low safety significance and has been entered into the
licensees corrective action program as CR 2008-001181, this violation is being treated
as an NCV, consistent with Section VI.A of the NRC Enforcement Policy: NCV 05000482/2008002-05, Failure to Follow Procedure.
2OS2 ALARA Planning and Controls (71121.02)
Inspection Planning
a. Inspection Scope
The inspectors assessed licensee performance with respect to maintaining individual
and collective radiation exposures ALARA. The inspectors used the requirements in
10 CFR Part 20 and the licensees procedures required by Technical Specifications as
criteria for determining compliance. The inspector interviewed licensee personnel and
reviewed:
- Outage or on-line maintenance work activities scheduled during the inspection
period and associated work activity exposure estimates, which were likely to
result in the highest personnel collective exposures
- Site-specific ALARA procedures
- Integration of ALARA requirements into work procedure and radiation work
permit documents
- Workers use of the low-dose waiting areas
- Radiation worker and radiation protection technician performance during work
activities in radiation areas, airborne radioactivity areas, or high radiation areas
- Self-assessments, audits, and special reports related to the ALARA program
since the last inspection
- Resolution through the corrective action process of problems identified through
postjob reviews and postoutage ALARA report critiques
- Corrective action documents related to the ALARA program and followup
activities, such as initial problem identification, characterization, and tracking
- 27 - Enclosure
- Effectiveness of self-assessment activities with respect to identifying and
addressing repetitive deficiencies or significant individual deficiencies
Documents reviewed by the inspector are listed in the attachment.
The inspector completed 5 of the required 15 samples and 4 of the optional samples.
b. Findings
No findings of significance were identified.
4. OTHER ACTIVITIES
4OA1 Performance Indicator Verification (71151)
.1 Data Submission
a. Inspection Scope
The inspectors performed a review of the data submitted by the licensee for the 4th
quarter 2007, performance indicators for any obvious inconsistencies prior to its public
release in accordance with IMC 0608, Performance Indicator Program.
This review was performed as part of the inspectors normal plant status activities and,
as such, did not constitute a separate inspection sample.
b. Findings
No findings of significance were identified.
.2 Unplanned Scrams per 7000 Critical Hours
a. Inspection Scope
The inspectors sampled licensee submittals for the unplanned scrams per 7000 critical
hours performance indicator for the period from the 4th quarter 2006 through the 4th
quarter 2007. To determine the accuracy of the performance indicator data reported
during those periods, performance indicator definitions and guidance contained in
Revision 5 of the NEI Document 99-02, Regulatory Assessment Performance Indicator
Guideline, were used. The inspectors reviewed the licensees operator narrative logs,
issue reports, event reports and NRC Inspection reports for the period of January 1,
2006, through December 31, 2007, to validate the accuracy of the submittals. The
inspectors also reviewed the licensees issue report database to determine if any
problems had been identified with the performance indicator data collected or
transmitted for this indicator and none were identified
Documents reviewed by the inspectors are listed in the attachment.
The inspectors completed one sample.
- 28 - Enclosure
b. Findings
No findings of significance were identified.
.3 Unplanned Scrams with Complications
a. Inspection Scope
The inspectors sampled licensee submittals for the unplanned scrams with
complications performance indicator for the period from the 4th quarter 2006 through the
4th quarter 2007. To determine the accuracy of the performance indicator data reported
during those periods, performance indicator definitions and guidance contained in
Revision 5 of the NEI Document 99 02, Regulatory Assessment Performance Indicator
Guideline, were used. The inspectors reviewed the licensees operator narrative logs,
issue reports, event reports and NRC Integrated Inspection reports for the period of
January 1, 2006, through December 31, 2007, to validate the accuracy of the submittals.
The inspectors also reviewed the licensees issue report database to determine if any
problems had been identified with the performance indicator data collected or
transmitted for this indicator and none were identified.
Documents reviewed by the inspectors are listed in the attachment.
The inspectors completed one sample.
b. Findings
No findings of significance were identified.
.4 Unplanned Transients per 7000 Critical Hours
a. Inspection Scope
The inspectors sampled licensee submittals for the unplanned transients per
7000 critical hours performance indicator for the period from the 4th quarter 2006
through the 4th quarter 2007. To determine the accuracy of the performance indicator
data reported during those periods, performance indicator definitions and guidance
contained in Revision 5 of the NEI Document 99 02, Regulatory Assessment
Performance Indicator Guidelines, were used. The inspectors reviewed the licensees
operator narrative logs, issue reports, maintenance rule records, event reports and NRC
integrated inspection reports for the period of January 1, 2006, through December 31,
2007, to validate the accuracy of the submittals. The inspectors also reviewed the
licensees issue report database to determine if any problems had been identified with
the performance indicator data collected or transmitted for this indicator and none were
identified.
Documents reviewed by the inspectors are listed in the attachment.
The inspectors completed one sample.
- 29 - Enclosure
b. Findings
No findings of significance were identified.
.5 Occupational Radiation Safety
a. Inspection Scope
The inspectors reviewed licensee documents for occupational exposure control
effectiveness from July 1 through December 31, 2007. The review included corrective
action documentation that identified occurrences in locked high radiation areas (as
defined in the licensees technical specifications), very high radiation areas (as defined
in 10 CFR 20.1003), and unplanned personnel exposures (as defined in NEI 99-02,
"Regulatory Assessment Indicator Guideline," Revision 5). Additional records reviewed
included ALARA records and whole body counts of selected individual exposures. The
inspector interviewed licensee personnel that were accountable for collecting and
evaluating the performance indicator data. In addition, the inspector toured plant areas
to verify that high radiation, locked high radiation, and very high radiation areas were
properly controlled. Performance indicator definitions and guidance contained in
NEI 99-02, Revision 5, were used to verify the basis in reporting for each data element.
Documents reviewed by the inspectors are listed in the attachment.
The inspectors completed one sample.
b. Findings
No findings of significance were identified.
a. Inspection Scope
The inspectors reviewed licensee documents for Radiological Effluent Technical
specification/Offsite Dose Calculation Manual Radiological Effluent Occurrences from
July 1 through December 31, 2007. Licensee records reviewed included corrective
action documentation that identified occurrences for liquid or gaseous effluent releases
that exceeded performance indicator thresholds and those reported to the NRC. The
inspector interviewed licensee personnel that were accountable for collecting and
evaluating the performance indicator data. Performance indicator definitions and
guidance contained in NEI 99-02, Revision 5, were used to verify the basis in reporting
for each data element.
Documents reviewed by the inspectors are listed in the attachment.
The inspectors completed one sample.
- 30 - Enclosure
b. Findings
No findings of significance 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 Items Entered Into the Corrective Action Program
a. Inspection Scope
As part of the various baseline inspection procedures discussed in previous sections of
this report, the inspectors routinely reviewed issues during baseline inspection activities
and plant status reviews to verify that they were being entered into the licensees
corrective action program at an appropriate threshold, that adequate attention was being
given to timely corrective actions, and that adverse trends were identified and
addressed. Attributes reviewed included: the complete and accurate identification of the
problem; that timeliness was commensurate with the safety significance; that evaluation
and disposition of performance issues, generic implications, common causes,
contributing factors, root causes, extent of condition reviews, and previous occurrences
reviews were proper and adequate; and that the classification, prioritization, focus, and
timeliness of corrective actions were commensurate with safety and sufficient to prevent
recurrence of the issue. Minor issues entered into the licensees corrective action
program as a result of the inspectors observations are included in the attached list of
documents reviewed.
These routine reviews for the identification and resolution of problems did not constitute
any additional inspection samples. Instead, by procedure they were considered an
integral part of the inspections performed during the quarter and documented in
Section 1 of this report.
b. Findings
No findings of significance were identified.
.2 Daily Corrective Action Program Reviews
a. Inspection Scope
In order to assist with the identification of repetitive equipment failures and specific
human performance issues for follow-up, the inspectors performed a daily screening of
items entered into the licensees corrective action program. This review was
accomplished through inspections of the stations daily condition report packages.
These daily reviews were performed by procedure as part of the inspectors daily plant
status monitoring activities and, as such, did not constitute any separate inspection
samples.
- 31 - Enclosure
b. Findings
No findings of significance were identified.
.3 Selected Issue Follow-up Inspection
a. Inspection Scope
During a review of items entered in the licensees corrective action program, the
inspectors selected the corrective action report listed below for a more indepth review.
The inspectors considered the following during the review of the licensee's actions:
(1) complete and accurate identification of the problem in a timely manner; (2) evaluation
and disposition of operability/reportability issues; (3) consideration of extent of condition,
generic implications, common cause, and previous occurrences; (4) classification and
prioritization of the resolution of the problem; (5) identification of root and contributing
causes of the problem; (6) identification of corrective actions; and (7) completion of
corrective actions in a timely manner.
- March 10, 2008, CR 2008-000790, automatic voltage control affected by scaffold
construction
The above constitutes completion of one indepth problem identification and resolution
sample.
Documents reviewed by the inspectors are listed in the attachment.
The inspectors completed one sample.
b. Findings
No findings of significance were identified.
.4 Routine Review of Items Entered into the Corrective Action Program for Access Control
to Radiologically Significant Areas and ALARA Planning and Controls
a. Inspection Scope
As part of the various baseline inspection procedures discussed in previous sections of
this report, the inspectors routinely reviewed issues during baseline inspection activities
and plant status reviews to verify that they were being entered into the licensees
corrective action program at an appropriate threshold, that adequate attention was being
given to timely corrective actions, and that adverse trends were identified and
addressed. Attributes reviewed included: the complete and accurate identification of the
problem; that timeliness was commensurate with the safety significance; that evaluation
and disposition of performance issues, generic implications, common causes,
contributing factors, root causes, extent of condition reviews, and previous occurrences
reviews were proper and adequate; and that the classification, prioritization, focus, and
timeliness of corrective actions were commensurate with safety and sufficient to prevent
recurrence of the issue. Minor issues entered into the licensees corrective action
program as a result of the inspectors observations are included in the attached list of
documents reviewed.
- 32 - Enclosure
These routine reviews for the identification and resolution of problems did not constitute
any additional inspection samples. Instead, by procedure they were considered an
integral part of the inspections performed during the quarter.
b. Findings
No findings of significance were identified.
4OA3 Followup of Events and Notices of Enforcement Discretion (71153)
.1 Technical Specification 3.0.3 Plant Shutdown due to ECCS voiding
a. Inspection Scope
The inspectors responded to the control room on January 11, 2008, and reviewed:
(1) operator logs, plant computer data, and/or strip charts for the above listed event to
evaluate operator performance in coping with nonroutine events and transients;
(2) verified that operator actions were in accordance with the response required by plant
procedures and training; and (3) verified that the licensee has identified and
implemented appropriate corrective actions associated with personnel performance
problems that occurred during the event.
Documents reviewed by the inspectors are listed in the attachment.
The inspectors completed one sample.
b. Findings
No findings of significance were identified.
.2 Notice of Enforcement Discretion (NOED) 08-4-001: NOED for Wolf Creek Nuclear
Operating Corporation CCP A Room Cooler Leak on February 13, 2008
a. Inspection Scope
On February 13, 2008, EDG B was out of service for planned maintenance, also one
offsite power source was out of service for I&C testing on the Train B degraded voltage
relays. On February 13, 2008, at 2:20 p.m., the Wolf Creek control room received a
report of a water leak from the room cooler for CCP A. At 3:50 p.m. on February 13,
2008, a circumferential flaw on an H-bend was discovered in SGL12A that resulted in
the NOED request. The inspectors reviewed the compensatory actions described in the
NOED. The inspectors observed the just-in-time training for the reactor operators which
consisted of the key operator actions that required a higher degree of assurance for
success to mitigate the NOED risk. Inspectors reviewed the offsite power surveillances,
the Sharpe station availability rounds, and the protected equipment signs.
Documents reviewed by the inspectors are listed in the attachment.
The inspectors completed one sample.
- 33 - Enclosure
b. Findings
.1 The inspectors questioned two operators regarding the just-in-time training for the most
risk significant reactor operator manual actions as shown in the Wolf Creek risk analysis.
The inspectors found that the operators had difficulty recalling the training objectives.
Subsequently, Wolf Creek re-briefed the control room crew on those manual actions.
Because this deficiency with the compensatory actions was resolved at approximately
the same time (within minutes) of the expiration of the 4-hour allowed outage time, and
before the Technical Specification requirement to be in Mode 3 within the subsequent
6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, the inspectors judged the deficiency to be minor.
.2 February 13, 2008, failure to establish reasonable expectation of operability
Introduction: A Green NRC identified NCV of TS 5.4.1 for failure to follow the operability
process on discovery of the CCP A room cooler leak.
Description: On February 13, 2008, EDG B was out of service for planned maintenance,
also one offsite power source was out of service for I&C testing on the Train B degraded
voltage relays. On February 13, at 2:20 p.m., the Wolf Creek control room received a
report of a water leak from the room cooler for CCP A. At 2:20 p.m., it could not be
established if the leak would cause a loss of structural integrity of the ESW system. Wolf
Creek Procedure AP 26C-004, Technical Specification Operability, Step 6.2.1 requires
continued operability decisions be made in the shift managers log. Wolf Creek made no
log entries at 2:20 p.m. stating the basis for immediate operability. At 3:50 p.m. Wolf
Creek control room logs state that CCP A had a room cooler leak and structural integrity
cannot be verified. Subsequent entry into TS 3.7.8 for the Pump ESW A caused EDG A
to be inoperable. TS 3.8.1, Condition I, states, that with three alternating current
sources inoperable, (both EDGs and on offsite source) TS 3.0.3 shall be entered. Wolf
Creek entered TS 3.0.3 at 3:50 p.m. and exited TS 3.0.3 at 4:13 p.m. when the inlet and
outlet valves to CCP As room cooler were closed. These log entries were after the fact
log entries made at approximately 5 p.m. to reflect the above sequence.
From interviews with control room operators on shift during this time, operators believed
that the most limiting TS action statement was TS 3.8.1.B.4.2.2 which is 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. This
was due to the fact that operators made an assumption that the leakage was not through
wall and that the cooler was operable prior to visual examination or other factual
information. The inspectors judged that, since structural integrity could not be assured
at 2:20 p.m., the room cooler was inoperable, as stated later in the Wolf Creek control
room logs. The inspectors could not locate any justification produced by Wolf Creek for
the room coolers operability after 2:20 p.m. In consultation with the Office of Nuclear
Reactor Regulation TS branch, the inspectors judged that it was not appropriate to make
such assumptions and wait for 1.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> to ascertain the nature of the leak when entry
into TS 3.0.3 would have been necessary and required action to be initiated within
1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> to place the unit in Mode 3 within 7 hours8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br />. Inspectors reviewed Part 9900,
Technical Guidance for Operability and for cases of ASME Code Class 3 leaks.
Part 9900 Technical Guidance states, in part, that an immediate operability declaration
shall be made with a reasonable expectation for continued operability within a period
commensurate with safety. During interviews, Wolf Creek staff stated that they had not
considered the extensive internal OE on through wall room cooler leaks during initial
operability reviews.
- 34 - Enclosure
Analysis: The inspectors determined that the failure to follow the operability process is a
performance deficiency. Traditional enforcement does not apply since there were no
actual safety consequences or potential for impacting the NRCs regulatory function, and
the finding was not the result of any willful violation of NRC requirements or Wolf Creek
procedures. The inspectors determined that this finding was more than minor because if
left uncorrected, it could become a more significant safety concern if the operability
procedures are not correctly applied. The inspectors evaluated the significance of this
finding using Phase 1 of IMC 0609, Appendix A, "Significance Determination of Reactor
Inspection Findings for At Power Situations," and determined that the finding screened to
Phase 2 because the finding represents an actual loss of safety function of a single train
of high head injection. A bounding risk of Green results from the Phase 2 presolved
worksheets. This result was obtained by using an exposure time of less than 3 days for
the scenario: Centrifugal Charging Pump PBG05A [Fails to Run]. The inspectors also
determined that the finding had crosscutting aspects in the human performance area
associated with decision making because the licensee failed to use conservative
assumptions in its operability decision and apply a requirement to demonstrate that the
room cooler is operable is in order to proceed rather than assuming that it is operable
with no supporting information H.1(b).
Enforcement: TS 5.4.1.a requires procedures be established, implemented, and
maintained covering the applicable procedures recommended in Regulatory Guide 1.33,
Appendix A. Appendix A, Section 1, recommends administrative procedures for safe
operation of the plant. Procedure AP 26C-004, Technical Specification Operability,
Revision 16 implements this requirement and states, in part, that continued operability
decisions shall be made in the shift managers log. Contrary to the above, on
February 13, 2008, at 2:20 p.m. CST, Wolf Creek did not implement its operability
procedure and establish operability for the CCP A room cooler. Because the finding is of
very low safety significance and has been entered into the corrective action program as
CR 2008-001647, this violation is being treated as an NCV, consistent with Section VI.A
of the NRC Enforcement Policy: NCV 05000482/2008002-06, Failure to Establish
Reasonable Expectation of Operability.
.3 Untimely Corrective Actions for CCP A Room Cooler Leads to NOED
Introduction: On February 13, 2008, the inspectors identified a noncited violation of
10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, for failure to take timely
corrective actions to prevent failure of the CCP A room cooler which resulted in the
NOED.
Description: The inspectors found that room Cooler SGL12A experienced leaks in
October 1999, May 2003, October 2003, August 2004, October 2006, and again in
February 2008. SGL12A was installed at the time of plant startup in 1985. On February
13, 2008, a circumferential flaw on an H-bend was discovered in SGL12A. Wolf Creek
subsequently initiated CR 2008-000467.
Problem Identification Reports (PIRs) 2005-2507 and 2004-0688 identified that leaks for
all room coolers had been an ongoing problem since at least April 2002. PIR 2004-0688
raised the eddy current minimum wall thickness acceptance criteria from 0-20 percent.
PIR 2005-2507 corrective actions had scheduled room Cooler SGL12A for replacement
on April 2, 2007, with a new stainless steel unit not susceptible to wall thinning leaks.
PIR 2005-2507 remains open for these corrective actions. On March 14, 2007,
- 35 - Enclosure
Wolf Creek chose to delay SGL12As replacement until Refueling Outage 16 due to the
required length of time to replace the cooler. SGL12A was then rescheduled for
replacement on March 22, 2008, the next refueling outage. The inspectors could not
locate an engineering evaluation to justify the replacement extension. During interviews,
Wolf Creek engineers stated that there is no formal failure analysis for the H-bend
failures. On February 13, 2008, SGL12A experienced its third H-bend through wall leak
and its sixth overall leak. The H-bend was then replaced as an interim measure.
Inspectors reviewed corrective action Procedure AP 28A-100, Condition Reports,
Revision 3 and found that a loss of a train to perform its safety function is considered a
significant deficiency requiring corrective action to prevent recurrence. The inspectors
reviewed PIRs 2005-2507 and 2004-0688, and CR 2008-0467 and found that Wolf
Creek designated each as nonsignificant which did not require actions to prevent
recurrence. Wolf Creek has subsequently implemented the corrective action identified in
PIR 2005-2507 to replace the SGL12A with a stainless steel unit during Refueling
Outage 16.
Analysis: The failure to take timely corrective actions was a performance deficiency.
Traditional enforcement does not apply since there were no actual safety consequences
or potential for impacting the NRC's regulatory function, and the finding was not the
result of any willful violation of NRC requirements or Wolf Creek procedures. The
inspectors determined that this finding was more than minor because it is associated
with the equipment performance attribute for the mitigating systems cornerstone; and, it
affected the cornerstone objective to ensure the availability, reliability, and capability of
systems that respond to initiating events to prevent undesirable consequences (i.e. core
damage). Specifically, this issue relates to the availability and reliability examples of the
equipment performance attribute because a failure mechanism was not corrected in
timely fashion and led to this failure. The inspectors evaluated the significance of this
finding using Phase 1 of IMC 0609, Appendix A, "Significance Determination of Reactor
Inspection Findings for At-Power Situations," and determined that the finding screened
to Phase 2 because the finding represents an actual loss of safety function of a single
train of high head injection, for greater than its Technical Specification 3.8.1.B.2 allowed
outage time of 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. Using an exposure time of less than 3 days for the scenario
Centrifugal Charging Pump PBG05A [Fails to Run], a bounding risk of Green results
from the Phase 2 presolved worksheets. Additionally, the cause of the finding has
crosscutting aspects in the human performance area associated with resources.
Specifically, Wolf Creek did not ensure adequate resources to maintain long-term plant
safety by minimizing the room coolers long-standing issues and preventive maintenance
deferrals H.2(a).
Enforcement: 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, requires,
in part, that for significant conditions adverse to quality, measures shall assure the cause
is determined and that actions are taken to preclude repetition. Corrective Action
Procedure AP 28-100, Condition Reports, Revision 3.states that a loss of a train to
perform its safety function is considered a significant deficiency requiring corrective
action to prevent recurrence. Contrary to the above, from October 23, 1999, to February
13, 2008, ECCS room Cooler SGL12A experienced multiple leaks. Specifically, the
licensee did not take corrective actions for approximately 9 years to prevent the
recurrence of leaks for Room Cooler SGL12A leading to the inoperability of a train of
ECCS equipment. This issue and the corrective actions are being tracked by Wolf
Creek in CR 2008-001673. Because the violation was of very low safety significance
and the issue was captured in the licensees corrective action program, this violation is
- 36 - Enclosure
being treated as an NCV consistent with Section VI.A of the NRC Enforcement Policy:
NCV 05000482/2008002-07, Untimely Corrective Actions for CCP Room Cooler Leads
.3 March 17, 2008, Reactor Trip due to XPB03 transformer trip
a. Inspection Scope
The inspectors responded to the control room on March 17, 2008, due to a reactor trip
from the XPB03 transformer trip, and reviewed: (1) operator logs, plant computer data,
and/or strip charts for the above listed event to evaluate operator performance in coping
with nonroutine events and transients; (2) verified that operator actions were in
accordance with the response required by plant procedures and training; and (3) verified
that the licensee has identified and implemented appropriate corrective actions
associated with personnel performance problems that occurred during the event. The
inspectors observed the reactor shutdown and cooldown.
Documents reviewed by the inspectors are listed in the attachment.
The inspectors completed one sample.
b. Findings
On March 17, 2008, plant operators observed that steam generator water level was
lowering and main feed pump speed was decreasing. Based on these indications, Wolf
Creek operators manually tripped the plant. Posttrip immediate actions and followup
actions were completed without deviation. An auto actuation of auxiliary feed water
occurred due to low/low steam generator water levels as expected but no other ECCS or
engineered safety feature actuations occurred. All plant equipment responded as
expected.
Following the trip, control room operators observed indications that the plant had
experienced a loss of the XPB03 13.8 kV to 4.16 kV nonsafety transformer which
powers PB003 4.16 kV nonsafety bus. Approximately 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> prior to the transformer
trip, Wolf Creek had removed from service XPB04 transformer for planned maintenance
and cross connected XPB04 transformer PB004 bus loads to the XPB03 transformer
PB003 bus. This arrangement powered all three condensate pumps from the PB003
4.16 kV bus. The PB003 bus powers condensate Pumps A and C and the PB004 bus
powers condensate Pump B. The XPB03 transformer trip resulted in losing power to all
three condensate pumps which tripped the main feed pumps on low suction pressure.
The licensees initial draft investigation of the cause of the transformer trip determined
that two phases of the XPB03 transformer 4.16 kV output cables had overheated and
failed. Additional investigation into the cable failures discovered that two multi-
directional conductor connectors used to terminate two phases of the 1000 million
circular mils (MCM) 4.16 kV bus cables were installed using the incorrect configuration.
The cable connector had been installed using a 1500-2000 MCM configuration which
resulted in the conductor connector bottoming out before applying sufficient compression
to ensure adequate connection to the cable.
- 37 - Enclosure
Pending completion of the licensees root cause determination and consequence
assessment by a Region IV Senior Reactor Analyst, additional inspection of the finding
is needed to determine significance. This issue is considered unresolved pending
additional NRC review of Wolf Creek root cause determination. This issue will be
tracked as: Unresolved Item (URI)05000483/2008002-08, Transformer Trip Resulted in
an Unplanned Reactor Trip and Forced Outage.
.4 (Closed) LER 05000482/2008-001-00, CCP A Room Cooler Out of Service Longer Than
Allowed Under Technical Specification 3.8.1.B.2
The inspectors reviewed LER 05000482/2008-001-00 to verify that the cause of the
Train A CCP exceeding its allowed outage time was identified and that corrective actions
were appropriate. See Section 4OA3.3 for additional information on the event and
enforcement actions taken. See also Notice of Enforcement Discretion for Wolf Creek
Nuclear Operating Corporation Regarding Wolf Creek Generating Station [TAC No
MD8098, NOED No. 08-4-001], under ADAMS Accession No. ML080520023 for more
information regarding the NOED. This LER is closed.
.5 (Closed) LER 05000482/2005-006-00: Unanalyzed Condition Related to Loss of
Reactor Coolant Pump Seal Cooling during a Postulated Appendix R Fire Event
Introduction. The inspectors identified an NCV of Technical Specification 5.4.1.d
because Procedure OFN-RP-017, "Control Room Evacuation," Revision 21, failed to
ensure that operators took the required actions to reestablish reactor coolant pump seal
cooling in a timely manner. Failure to establish seal cooling in a timely manner could
have resulted in a small break loss of coolant accident. This finding was determined to
be of very low risk significance (Green).
Description. While timing operator actions during a 2005 triennial fire protection
inspection (NRC Inspection Report 05000482/2005008, Section 1R05.6.b(2)), the
inspectors determined that control room operators could not reestablish seal cooling to
the reactor coolant pumps in a timely manner. The failure to reestablish seal cooling
within 21 minutes would degrade the seals and could result in a small break loss of
coolant accident. The delay in reestablishing seal cooling to the reactor coolant pumps
allows the seals to overheat and the subsequent flow of relatively cool water shatters the
seals and allows for excessive leakage. Specifically, the inspectors postulated circuit
failures that required operators to start the Train B EDG manually, as specified in
Procedure OFN-RP-017, Attachment C, Step 10, and manually open
Valve BN-LCV-112E, Train B CCP suction from the refueling water storage tank, as
specified in Attachment C, Step 24.
The licensee indicated that they had planned to revise Procedure OFN-RP-017 in
response to information contained in Information Notice 2005 14, "Fire Protection
Findings on Loss of Seal Cooling to Westinghouse Reactor Coolant Pumps," dated
June 5, 2005, and Westinghouse WCAP-16396 NP, "Reactor Coolant Pump Seal
Performance for Appendix R Assessments," dated January 2005. The licensee reported
that the NRC used a more conservative approach to develop the time line for
reestablishing seal cooling to the reactor coolant pumps than they had previously used.
The failure to ensure that operators could reestablish seal cooling to the reactor coolant
pumps within the prescribed time could cause failure of the pump seals and increase the
- 38 - Enclosure
leakage upon reestablishing the cooling such that pressurizer level would decrease
below the indicating range. The licensee documented this deficiency in their corrective
action program as PIR 2005-03209. The licensee modified Procedure OFN-RP-017 to
require operators to trip the reactor coolant pumps immediately.
The inspectors reviewed the physical configuration of the control room and verified that a
fire would have to affect two separate panels and disable specific components on the
panels. The control switch for the charging pump suction valve is located on
Panel RL001 and the control switch for the Train B EDG is located on Panel RL015.
The top of Panel RL015 opens to the ceiling (i.e. the floor of the upper cable spreading
room) although penetrations are sealed. The inspectors verified that approximately
3 feet separates the front of Panel RL015 from the rear of Panel RL001 and that 7 feet
separate the switches on the separate panels. Panel RL001 has a top vent that allows
heat to escape. Neither panel has front vents; consequently, air does not readily flow
through the panels. Channels separate the Trains A and B components within each
panel. Because of the channel separation within each panel, a high likelihood exists that
the Train A components would be available.
Analysis. This performance deficiency resulted from an inadequate postfire safe
shutdown procedure. The inspectors determined the finding is more than minor in that it
affected the ability to achieve and maintain hot shutdown following a control room fire.
This finding is associated with the mitigating systems cornerstone attribute of protection
against external factors (e.g. fire). This finding affected the mitigating systems
cornerstone objective to ensure the availability, reliability, and capability of systems that
respond to external events (such as fire) to prevent undesirable consequences.
Consequently, the inspectors evaluated these deficiencies using IMC 0609, Appendix F.
The inspectors determined that this procedure deficiency had more than minor impact on
the ability to implement the postfire safe shutdown procedure; consequently, the
inspectors assigned the issue a moderate degradation rating. The deficiency required a
Phase 3 evaluation since Appendix F did not explicitly apply to fires that result in
evacuating the control room.
The NRC senior reactor analyst assigned a generic fire ignition frequency for the control
room which was slightly higher than the value in the IPEEE for Severe Accident
Vulnerabilities. The analyst multiplied the fire ignition frequency by a severity factor and
a nonsuppression probability indicating that operators failed to extinguish the fire within
20 minutes assuming 2 minute detection that requires a control room evacuation. The
resulting evacuation frequency is:
Control Room Evacuation Frequency = fire ignition frequency for the control room *
severity factor * NP control room evacuation =
Control Room Evacuation Frequency = 1.09E-02/year * 0.1 * 1.30E-02 = 1.42E-05/year
The analyst estimated the probability of a fire induced failure as a two wire short and
determined this probability to be 0.6 squared for a resulting probability of 0.36. The
analyst calculated the resulting frequency of occurrence by multiplying the control room
evacuation frequency by the two wire short for a value of 5.10E 06/year.
The analyst determined the delta conditional core damage probability by subtracting the
base case conditional core damage probability (0.1) from the assumed fire damage
- 39 - Enclosure
conditional core damage probability (1.0) for a value of (0.9). The bounding delta
conditional core damage frequency for a 1 year exposure is the frequency of occurrence
(5.10E-06/year) multiplied by the delta conditional core damage probability (0.9) for a
value of 4.59E-06.
The analyst then qualitatively assessed the probability that the specific fires necessary
would occur. The fire had to affect components located in two physically separated
panels, as described below:
- On Panel RL015 the protected Train B diesel generator control power and the
Train A diesel generator control power prior to the transfer. Affecting both
components separated by 1.32 meters has a low likelihood.
- On Panel RL001 are the controls and valves for the centrifugal charging pumps
that provide seal cooling to the reactor coolant pumps. A distance of 3 feet
separated the front of Panel RL015 from the rear of Panel RL001; in addition, a
distance of 7 feet separated the switches on Panel RL001 and the switches on
Panel RL015.
The licensee installed fire resistant cables qualified in accordance with Institute of
Electrical and Electronics Engineers (IEEE) Standard 383-1974, "IEEE Standard for
Type Test of Class IE Electric Cables, Field Splices, and Connections for Nuclear Power
Generating Stations," throughout the plant.
The analyst referred to NUREG/CR-6850, "Fire PRA Methodology for Nuclear Power
Facilities," Section 11.5.2 and the test results described in NUREG/CR-4527, "An
Experimental Investigation of Internally Ignited Fires in Nuclear Power Plant Control
Cabinets: Parts 1 and 2," to characterize the effects of cabinet spacing. The analyst
determined that NUREG/CR-6850 discussed that the review of control room fires
determined that none of the fires affected components beyond the point of ignition and
that in all cases operators extinguished the fires with hand held extinguishers. The
testing results reported in NUREG/CR-4527, Section 4.2.2, indicated that fire growth
depended on ventilation flow through the cabinet to provide fresh oxygen and fire spread
to an adjacent cabinet is very dependent upon the location of the cabinet, the barriers
between the cabinets, and the qualification of the wires. The laboratory performed the
testing on adjacent cabinets with one inch separation and single and double walls. The
testing demonstrated that the worst-case spread of fire outside a cabinet occurred with
unqualified cables and only extended 0.5 meters.
Considering the distance between the cabinets of 1 meter and the use of qualified
cables, the analyst concluded that it would be highly unlikely for a fire to move from one
cabinet to another within the 20 minute period before operators suppressed the fire or
restored seal injection. Because of the separation, the analyst concluded that the
qualitative factors would reduce the bounding value such that this deficiency had very
low risk significance (Green). This finding did not have crosscutting aspects since the
performance deficiency occurred outside of the assessment period.
Enforcement. Technical Specification 5.4.1.d states the licensee will establish,
implement, and maintain procedures for implementing the fire protection program.
Procedure OFN-RP-017, "Control Room Evacuation," Revision 21, specified
requirements to reestablish seal cooling to the reactor coolant pumps. Contrary to the
- 40 - Enclosure
above, the inspectors determined that operators could not implement the steps of
Procedure OFN-RP-017 within the critical time to prevent seal damage, which would
result in a small break loss of coolant accident. Because this finding is of very low safety
significance and the licensee entered the deficiency into the corrective action program,
the inspectors considered this issue as a NCV, consistent with Section VI.A of the NRC
Enforcement Policy: NCV 05000482/2008002-09, Failure to Reestablish Timely Seal
Cooling for the Reactor Coolant Pumps.
.6 (Closed) LER 05000482/2005-007-00: Unanalyzed Condition Related to Loss of EDG
Field Flashing during an Appendix R Fire Event
Introduction: The inspectors documented the enforcement related to this LER in
Section 4OA7. This LER is closed.
Description: Because of the 2005 NRC triennial fire protection inspection, the licensee
reviewed actions specified in Procedure OFN-RP-017 to ensure operators could
implement the actions in the time specified. During review of the procedure, the licensee
evaluated the EDG start circuits to determine if a control room fire affected their
operability. From review of the circuits, the licensee determined that the automatic start
circuits remained unaffected. However, while reviewing circuits associated with field
flashing, the licensee determined that control circuit fuses could blow if the fire causes a
short to ground in certain cables and that the loss of control power will prevent field
flashing.
As immediate corrective actions, the licensee staged replacement fuses for each diesel
generator, added steps in Procedure OFN-RP-017 directing the use of the fuses for a
field flash circuit failure, and initiated PIR 2005-3333.
Analysis: The performance deficiency associated with this finding involved failure to
have an adequate postfire safe shutdown procedure for response to a control room fire.
This finding is more than minor because it is associated with mitigating systems
cornerstone attribute of protection from external factors (fire) and affects the associated
cornerstone objective to ensure the availability, reliability, and capability of systems that
respond to initiating events to prevent undesirable consequences.
On Panel RL015 the licensee had separated the hand switches for the Trains A and B
EDG by 52 inches (~1.32 m). A fire affecting the hand switches could fail the
corresponding field flash relay fuse locally and render the affected EDG(s) inoperable.
Further, a fire in Panel RL015 could cause a loss of offsite power. The licensee assigns
the Train B as the safe shutdown path for a control room fire and does not credit any
Train A components.
The IPEEE assigns a fire frequency of 9.45E-05/yr for a single control room panel. To
bound this assessment, the analyst assumed that fires in adjacent cabinets could spread
one cabinet over; therefore, the analyst increased the fire frequency by a factor of 3 to
2.84E-04/yr. Using NUREG/CR-6850, the analyst estimated the risk of losing both
Trains A and Train B EDG. Specifically, using NUREG/CR-6850, Appendix L, "Appendix
for Chapter 11, Main Control Board Fires," Figure L-1, the analyst determined the
likelihood of disabling both hand switches separated by 1.32 meters. The value
determined from the figure accounted for the nonsuppression probability and the severity
factor. Multiplying the likelihood value of 1.00E-03 resulted in a fire frequency affecting
- 41 - Enclosure
both emergency diesel generators as 2.84E-07/yr. Fires originating in other locations
would not result in a change to the risk significance of the finding.
The analyst made a bounding assumption that fire damage to the Train B EDG field
flash circuit would not be recovered and that the unprotected Train A EDG would also be
lost. However, in the base case (without the performance deficiency), the analyst
assumed the Train B EDG would always be recovered. The TDAFW pump will fail upon
loss of direct current power. This leaves only the recovery of offsite power as a means
to avoid core damage. In general, the actions to restore offsite power would entail very
simple breaker manipulations and it is likely that at least 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> would be available,
except in the rare cases where the TDAFW pump fails.
Using the SPAR-H human performance method, the analyst applied full credit for
diagnosis and computed an action human reliability analysis value of 2E-02 for the
short-term sequence associated with TDAFW failure and 2E-03 TDAFW success. Both
assume high stress and the available time accounts for the order of magnitude
difference. The Wolf Creek SPAR model assigns an overall probability of 2.2E-02 that
the TDAFW pump will not be available for mitigation. This results in the following two
sequences that comprise the bounding estimate of the delta core damage frequency (the
exposure period of the finding is 1 year):
Auxiliary feedwater unavailable: (2.84E-07/yr) (2.2E-02) (2E-02) = 1.25E-10
Auxiliary feedwater success: (2.84E-07/yr) (2E-03) = 5.64E-10
The analyst determined a bounding risk estimate of 6.89E-10/yr. and has minimal affect
on large early release frequencies. Therefore, the analyst concluded this issue had very
low risk significance (Green). The inspectors determined this finding had no crosscutting
aspect since it did not reflect current licensee performance.
Enforcement: The inspectors documented this licensee identified violation in
Section 4OA7.
.7 (Closed) LER 05000482/2006-001-00: Potential for Fire Induced Damage to Motor
Operated Valves during an Appendix R Fire Event
This licensee initiated this LER to document that a control room fire could affect
40 motor-operated valves. This LER described the same issue as
URI 05000482/2005008-06, "Failure to Evaluate Adequately Fire Protection Program
Deficiencies," which was closed in Section 4OA5.2. This LER is closed.
.8 (Closed) LER 05000482/2006-002-00: Potential for Fire Induced Damage to Class 1E
Electrical Equipment Air Conditioning Units during an Appendix R Event
On May 24, 2006, while performing a postfire safe shutdown review for Fire Area C-35,
the licensee discovered an unanalyzed condition. A fire in this area could prevent
operation of both Class 1E electrical equipment air conditioning units if a fire damaged
the automatic fire isolation circuit on the fan units. The loss of the Class 1E air
conditioning units would not directly result in loss of capability to shut down the facility
safely. Rather, room heating beyond design limits could reduce the life of electrical
components within the switchgear.
- 42 - Enclosure
As immediate corrective actions, the licensee established a continuous fire watch for
Fire Area C-35, initiated a temporary change to Procedure OFN-KC-016, "Fire
Response," and staged jumpers. The licensee included this deficiency in their corrective
action program as CR 2006-000551. Long-term corrective actions involved installing a
bypass switch on Panel RP068.
Disposition of this LER is in Section 4OA7. This LER is closed.
4OA5 Other Activities
.1 (Closed) Apparent Violation 05000482/2005008-05: Inadequate Alternative Shutdown
Procedure
The issue documented by this apparent violation is the same issue discussed in
LER 05000482/2005-006-00, "Unanalyzed Condition Related to Loss of RCP Seal
Cooling during a Postulated Appendix R Fire Event," in Section 4OA3.5. The inspectors
discussed the enforcement for this event in Section 4OA3.5. This apparent violation is
closed.
.2 (Closed) Unresolved Item 05000482/2005008-06: Failure to Adequately Evaluate Fire
Protection Program Deficiencies
Documents reviewed by the inspectors are listed in the attachment.
Introduction. The inspectors identified an NCV of License Condition 2.c(5) because the
licensee failed to evaluate the impact of a motor operated valve failure mechanism on
their ability to implement postfire safe shutdown following a control room evacuation.
The licensee determined that the failure mechanism affected 38 motor-operated valves
and upon failure could affect their ability to implement their postfire safe shutdown
procedure. This finding was determined to be of very low risk significance (Green).
Description. During a triennial fire inspection in 2005 (NRC Inspection
Report 5000482/2005008), the inspectors determined that the licensee had not
effectively reviewed industry operating experience information on two previous
occasions. Consequently, the licensee failed to determine the population of motor
operated valves that would be susceptible to mechanistic damage. The damage could
result if fire induced short circuits bypassed the torque and limit switches. The
inspectors identified four valves that could have had their protection bypassed and
operators would need to operate them following a control room fire, as specified in
Procedure OFN-RP-017.
The NRC issued Information Notice 92-18, "Potential for Loss of Remote Shutdown
Capability during a Control Room Fire," which described conditions related to a control
room fire that causes operators to evacuate the control room. Specifically, a fire in the
control room could cause hot short circuits between control wiring and power sources for
motor-operated valves needed for safe shutdown and operated from remote locations.
However, hot short circuits combined with the absence of thermal overload, torque
switch and limit switch protection, could cause valve damage before the operator shifted
control of the valves to the remote shutdown panel.
- 43 - Enclosure
The licensee identified 38 Train B motor-operated valves potentially affected and
initiated PIR 2005-3314 to resolve this deficiency. The licensee developed a
modification that altered the control circuit for each valve to prevent a control room fire
from bypassing the torque/limit switches or failing the thermal overload.
During this inspection, the inspectors verified that the motor-operated valves resided on
five control panels. The inspectors evaluated the physical separation of the safety
related postfire safe shutdown train from the opposite safety-related train controls and
the separation among the safety related and nonsafety-related controls. The inspectors
also considered remaining capability from other systems on separate panels. Functions
related to postfire safe shutdown needed to achieve and maintain hot shutdown were
located on four control room panels. Specifically, the motor-operated valves could affect
the following functions on the listed panels:
- Panel RL001 - charging/letdown and seal injection flow to the reactor coolant
pumps,
- Panel RL005 - auxiliary feedwater flow suction valves from the condensate
storage tank and the essential service water system,
- Panel RL017 - residual heat removal valves needed to achieve cold shutdown,
- Panel RL018 - boron injection valves used in maintaining pressurizer level, and
- Panel RL019 - essential service water and component cooling water valves.
The limiting valves on this panel to cause a loss of function involve the control
switches for the critical loop discharge and return valves for both component
cooling water safety related trains.
The inspectors confirmed that the licensee used cables with the following characteristics;
- The licensee utilized IEEE-383 qualified wire insulation and cable jackets.
- The valves had seven conductor cable wiring that required a smart hot short from
one conductor to the other.
- The valves had control power transformers.
Analysis. The inspectors determined this was a performance deficiency because the
licensee failed to ensure that components necessary to safely shutdown the reactor
would remain operable following a fire. This deficiency was more than minor in that it
had the potential to impact the mitigating systems cornerstone objective to ensure the
availability, reliability, and capability of systems that respond to external events (such as
fire) to prevent undesirable consequences.
The NRC senior reactor analyst assigned a generic fire ignition frequency for the control
room (FIFCR), which was slightly higher than the value in the IPEEE. The analyst
multiplied the fire ignition frequency by a severity factor and a nonsuppression
probability indicating that operators failed to extinguish the fire within 20 minutes
assuming 2 minute detection that requires a control room evacuation (NPCRE). The
resulting evacuation frequency is:
- 44 - Enclosure
Control Room Evacuation Frequency = FIFCR * SF * NPCRE =
Control Room Evacuation Frequency = 1.09E-02/year * 0.1 * 1.30E-02 = 1.42E-05/year
The analyst estimated the probability of a fire induced failure as a two wire short and
determined this probability to be 0.6 squared for a resulting probability of 0.36. The
analyst calculated the resulting frequency of occurrence by multiplying the control room
evacuation frequency by the two wire short for a value of 5.10E-06/year.
The control room had 103 panels with the wiring and circuits for the affected valves
residing in five panels. Therefore, the probability that a control room fire would affect the
panels of interest is 4.85E-02. The resulting mitigation frequency is the frequency of
occurrence multiplied by the partial fraction represented by the affected cabinets for a
value of 2.47E-07.
Given that the change in core damage frequency would be determined by multiplying the
mitigation frequency value determined above by a conditional core damage probability
equal to or less than one, the analyst determined this deficiency had very low risk
significance (Green). This finding did not have crosscutting aspects since the
performance deficiency occurred outside of the assessment period.
Enforcement. License Condition 2.c(5) states that the licensee shall maintain in effect all
provisions of the approved fire protection program as described in the licensee's USAR.
The USAR, Appendix 9.5A, Table 9.5a-1, Section C.8 states that the licensee will
promptly identify and correct deficiencies that affect fire protection. 10 CFR Part 50.48,
requires all plants to meet Appendix R,Section III.G.Section III.G.1.a requires that one
train of safe shutdown equipment be capable of achieving and maintaining hot shutdown
conditions from either the control room or the emergency control station(s) and shall be
free of fire damage. Contrary to the above, the inspectors determined that the licensee
failed to ensure that, following a control room fire, operators would be able to manipulate
postfire safe shutdown motor-operated valves because of damage caused by fire.
Because the licensee included this deficiency in their corrective action program and
because the deficiency had very low safety significance, the inspectors considered this
issue as an NCV, consistent with Section VI.A of the NRC Enforcement Policy: NCV 05000482/2008002-10, Failure to Analyze Motor-Operated Valve Circuits.
4OA6 Meetings, Including Exit
On February 20, 2008, the inspectors presented the results of the fire protection
inspection open item review and closeout to Mr. L. Ratzlaff, Manager, Support
Engineering, and other members of licensee management. The licensee acknowledged
the information presented.
On April 4, 2008, the inspectors presented the occupational radiation safety inspection
results to Mr. M. Sunseri and other members of his staff who acknowledged the findings.
The inspector confirmed that proprietary information was not provided or examined
during the inspection.
On April 11, 2008, the resident inspectors presented the inspection results of the
resident inspections to Mr. S. Hedges, Vice President Oversight, and other members of
the licensee's management staff. The licensee acknowledged the findings presented.
- 45 - Enclosure
The inspectors noted that while proprietary information was reviewed, none would be
included in this report.
4OA7 Licensee-Identified Violations
The following violations of very low significance (Green) were identified by the licensee
and are violations of NRC requirements which meet the criteria of Section VI of the NRC
Enforcement Policy, NUREG-1600, for being dispositioned NCVs.
- Licensee Technical Specification 5.7.1.b states in part that access to high
radiation areas with dose rates not exceeding 1.0 Rem/hour at 30 centimeters
from the radiation source shall be controlled by means of a radiation work permit
that includes specification of radiation dose rates in the immediate work area and
other appropriate radiation protection equipment and measures. Contrary to
these regulations, on January 13, 2008, two quality control inspectors entered a
pipe chase, a posted high radiation area, on the 1988 elevation of the auxiliary
building using the wrong radiation work permit. The radiation work permit used
by the licensee inspectors did not allow entry into a high radiation area. The
violation was entered into the licensees corrective action program as
CR 2008-00112. The finding was determined to be of very low safety
significance because it did not involve: (1) ALARA planning and controls, (2) an
overexposure, (3) a substantial potential for overexposure, or (4) an impaired
ability to assess dose.
- Technical Specification 5.4.1.d specified that the licensee have fire protection
procedures established, maintained, and implemented. Procedure OFN-RP-017,
"Control Room Evacuation," Revision 21, specified actions for a fire in the control
room. Contrary to this requirement, the licensee determined that the procedure
failed to provide mitigating actions for a failure of the field flash relay control
circuit because of possible fire damage. As described in Section 4OA3.6, this
finding was of very low safety significance.
- Title 10 of the Code of Federal Regulations, 10 CFR 50.48, requires all plants to
meet Appendix R,Section III.G. Appendix R,Section III.G.2, specified that for
equipment and cables of redundant trains of systems necessary to achieve and
maintain hot shutdown located within the same fire area outside of primary
containment shall be separated by one of the means specified or a diverse
means implemented. Contrary to this requirement, the licensee did not provide
the required separation and had not implemented a diverse means to ensure the
required Class 1E air conditioning units would remain functional. This finding
had a low degradation rating because of the very low likelihood of occurrence
and the ability to achieve safe shutdown did not become directly affected;
consequently, the deficiency had very low safety significance. The licensee
included this item in their corrective action program (refer to Section 4OA3.8)
- 46 - Enclosure
SUPPLEMENTAL INFORMATION
- KEY POINTS OF CONTACT
Licensee
R. A. Muench, President and Chief Executive Officer
M. Sunseri, Vice President Operations and Plant Manager
S. E. Hedges, Vice President Oversight
K. Scherich, Director Engineering
T. East, Manager, Emergency Planning
P. Bedgood, Superintendent, Chemistry/Radiation Protection
ITEMS OPENED, CLOSED, AND DISCUSSED
Opened
05000482/2008002-03 URI Containment sump net positive suction head losses.
(Section 1R15)05000482/2008002-08 URI Transformer trip resulted in an unplanned reactor trip and
forced outage (Section 4OA3.3)
Opened and Closed
05000482/2008002-01 NCV Failure to implement fire protection impairment control
permit requirements and compensatory measures.
(Section 1R05)05000482/2008002-02 NCV Performing prohibited elective maintenance on offsite
power during EDG maintenance. (Section 1R13)05000482/2008002-04 NCV Failure to control area as a locked high radiation area.
(Section 2SO1(1))05000482/2008002-05 NCV Failure to follow Procedure. (Section 2SO1(2))05000482/2008002-06 NCV Failure to establish reasonable expectation of operability
(Section 4OA3.2(2))05000482/2008002-07 NCV Untimely corrective actions for CCP room cooler leads to
NOED. (Section 4OA3.2(3))05000482/2008002-09 NCV Failure to reestablish timely seal cooling for the reactor
coolant pumps (Section 4OA3.5)
-1- Attachment
05000482/2008002-10 NCV Failure to analyze motor-operated valve circuits
(Section 4OA5.2)
Closed
05000482/2008-001-00 LER CCP A Room Cooler Out of Service Longer Than Allowed
Under Technical Specification 3.8.1.B.2 (Section 4OA3.4)
05000482/2005-006-00 LER Unanalyzed Condition Related to Loss of Reactor Coolant
Pump Seal Cooling during a Postulated Appendix R Fire
Event (Section 4OA3.5)
05000482/2005-007-00 LER Unanalyzed Condition Related to Loss of EDG Field
Flashing during an Appendix R Fire Event
(Section 4OA3.6)
05000482/2006-001-00 LER Potential for Fire Induced Damage to Motor Operated
Valves during an Appendix R Fire Event (Section 4OA3.7)
05000482/2006-002-00 LER Potential for Fire Induced Damage to Class 1E Electrical
Equipment Air Conditioning Units during an Appendix R
Event (Section 4OA3.8)05000482/2005008-05 AV Inadequate Alternative Shutdown Procedure
(Section 4OA5.1)05000482/2005008-06 URI Failure to Adequately Evaluate Fire Protection Program
Deficiencies (Section 4OA5.2)
LIST OF DOCUMENTS REVIEWED
In addition to the documents referred to in the inspection report, the following documents were
selected and reviewed by the inspectors to accomplish the objectives and scope of the
inspection and to support any findings:
Section 1R01: Adverse Weather
Procedures
STN EF-020B, ESW Train B Warming Line Verification, Revision 6
SYS EF-205, ESW/CIRC Water Cold Weather Operations, Revision 19
AI 14-006, Severe Weather, Revision 7
Section 1R04: Equipment Alignment
Procedures
CKL EF-120, Essential Service Water Valve, Breaker and Switch Lineup, Revision 41
SYS KJ-121,Diesel Generator Lineup for Auto Ops, Revision 39
-2- Attachment
Work Order
06-289610-000
Work Request
07-063628
Miscellaneous
Engineering Disposition, Relocate I/P From The ARVs, ABPV001 Thru 004, Revision 6
Wolf Creek Generating Station USAR, Revision 19
Section 1R05: Fire Protection
Procedures
ALR KC-888, Fire Protection Panel KC-008 Alarm Response, Revision 15
AP 10-106, Fire Preplans, Revision 5
OFN ST-003, Natural Events, Revision 13A
STN FP-815A, Heat Trip Actuation Device Operational Test Zones BZ 503,
016/SZ1-5Z47,1-2Z28, A Train Emergency Diesel Generator and ESF Transformer,
Revision 3
Condition Report
2007-002929
Work Request
07-063647
Work Orders
06-284430-000 06-284436-000
Drawings
E-OFO221, Fire Detection/Protection System-Yard Transformer Area EL. 2000'-0", Revision 5
M-13EA01, Piping Orthographic Service Water System Communication Corridor, Revision 6
M-13EF01, Piping Isometric Essential Service Water System Control Bldg. A & B Train ,
Revision 11
-3- Attachment
Miscellaneous
Wolf Creek Generating Station Individual Plant Examination Summary Report, September 1992
Post Fire Safe Shutdown Area Analysis, E-1F9910, Revision 2
Fire Hazard Analysis Fire Area H-1, Revision 0
Prefire Plan, Auxiliary Building Prefire Plans, Revision 6
Prefire Plan, Fire Protection Water Supply and Hydrant Locations, Revision 0
Fire Hazard Analysis, Fire Area CST & RWST, Revision 0
Section 1R11: Operator Requalification
Procedures
AI 21-100, Operations Guidance and Expectations, Revision 8
AP 21-001, Conduct of OPS, Revision 36A
APF 06-02-001, Emergency Action Levels, Revision 8
EDI 23M-050, Monitoring Performance to Criteria and Goals, Revision 3
EPP 06-06, Protective Action Recommendations, Revision 4
Miscellaneous
Operations Requalification Cycle 07-01, Revision 0
Section 1R12: Maintenance Effectiveness
Performance Improvement Requests
2007-1952 2007-1953 2007-2100 2007-2141
96-2671
Work Requests
07-061766 07-061883 07-061884 07-060117
07-060141 07-060514 07-059846
Work Orders
07-298545-000 07-296463-000 07-292308-000 07-291903-000
07-291889-000 07-301051-001 07-301051-011 07-293935-000
07-293935-003 07-294968-003 07-294968-000 07-295395-000
07-295396-000 05-270547-001 06-287445-000 05-271470-000
-4- Attachment
Condition Reports
2007-000860 2007-000879 2007-000897 2007-000943
2007-000988 2007-004154
Maintenance Rule
Maintenance Rule Scoping Evaluation for System BB - Reactor Coolant System
Maintenance Rule Scoping Evaluation for System INS -Reg. Guide 1.97 Instrumentation
Maintenance Rule Final Scoping Evaluation AB-05
Maintenance Rule Final Scope Evaluation GN-01
Maintenance Rule Final Scope Evaluation GN-02
Maintenance Rule Final Scope Evaluation GN-03
Maintenance Rule Final Scope Evaluation GN-04
Maintenance Rule Final Scope Evaluation GN-06
Maintenance Rule Final Scope Evaluation GN-08
Maintenance Rule Final Scope Evaluation KA-01
Maintenance Rule Final Scope Evaluation KA-03
Maintenance Rule Final Scope Evaluation KA-04
Maintenance Rule Final Scope Evaluation KA-06
Miscellaneous
EDI 23M-050 Attachment B, Functional Failure Determination Checklist
M-12KA01, Piping & Instrumentation Diagram Compressed Air System, Revision 27
INC C-1000, Calibration of Miscellaneous Components, Revision 7
STS AB-201A, Main Steam Isolation Bypass Inservice Valve Test, Revision 14
Calculation E-11005, List of Loads Supplied by Emergency Diesel Generator, Revision 32
BD-EMG ES-04, Natural Circulation Cooldown, Revision 8
Engineering Disposition 116451-10
USAR 1.2.9.6, Compressed Air Systems
EDI 23M-050, Engineering Desktop Instruction Monitoring Performance to Criteria Goals,
Revision 3
Calculation AN-99-031, Development of PSA based Reliability Performance Criteria for
Maintenance Rule, Revision 0
-5- Attachment
Section 1R13: Maintenance Risk Assessments and Emergent Work Control
Procedures
AIF 22C-006-01, Checklist for Emergent Work, Revision 4, September 2, 2007
AP 16E-002, Post Maintenance Testing Development, Revision 6A
AP 22C-003, Operational Risk Assessment Program, Revision 11
APF 21-001-02, Control Room Turnover Checklist, Revision 23, September 2-4, 2007
APF 21-001-06, Site Operator Relief Checklist, Revision 5, September 2-4, 2007
APF 22C-003-001, Operational Risk Assessment, Revision 0, September 2-4, 2007
MPE RC-001, Room Cooler Maintenance, Revision 8B
PSA-05-0020, WCGS PRA Basic Event Data Files, Appendix E, Revision 1
STN AB-003, Main Steam Iso Vlv Acc Discharge, Revision 11
Condition Reports
2007-004045 2007-004056 2007-004098 2007-004084
2007-004075
Work Orders
07-300584-000 07-300584-001 07-300584-002 07-292506-000
06-282700-000 07-297055-000
Miscellaneous
Operational Risk Assessment, Schedule Week 404
Commercial Grade Dedication Evaluation Number 021-E-0003
Calculation GL-M-002, Calculate tube plugging allowance for Aerofin (Cu-Ni) coils for --
Electrical Penetration Room Coolers (SGL15A & SGL15B), Revision 0
M-612-010-2, 39E Air Handling Units, Revision 2
USAR Figure 6.2.1-80, Main Steam Line Break Analysis, Case 9, Containment Temperature,
Revision 6
Design Specification for Room Coolers for the Wolf Creek Generating Station, Revision 9
Procedures
Section 1R15: Operability Evaluations
Procedures
CKL KA-121, Instrument Air Valve Lineup, Revision 9A
CKL NT-120, Nitrogen System Normal Valve Lineup, Revision 23
MPEE009Q-03, Inspection and testing of Siemens vacuum circuit breakers
-6- Attachment
STS BG-002, ECCS Valve Check and System Vent, Revision 25
STS KJ-001A , Integrated D/G and Safeguards Actuation test, Train A performed on
November 12, 2003
Condition Reports
2007-004329 2007-003704 2007-003462
Work Orders
03-253210, 03-25931 and 01-224513 demonstrating verification of charging spring times for
selected Siemens vacuum circuit breakers
Miscellaneous
Drawing E-11005, List of loads supplied by EDG
PIR 2003-3463, CCW pump breaker design issue
Technical Requirements Manual 3.4.17, Structural Integrity
Technical Requirements Manual Bases 3.4.17, Structural Integrity
M-13KA47, Small Piping Isometric Nitrogen Back-Up Gas Supply Auxiliary BLDG., Revision 8
M-13KA46, Small Piping Isometric N2 Back-Up Gas Supply Auxiliary BLDG. & Turbine BLDG.,
Revision 9
M-13KA51, Small Piping Isometric N2 Back-Up Gas Supply Auxiliary Building, Revision 1
D-79-600, 25 ft3 Gas Accumulator Bechtel Power Company (SNUPPS), Revision 5
OP EVAL Evaluation of as found voids in ECCS suction piping
Section 1R18: Plant Modifications
Procedure
Procedure AP 29B-002, ASME Code Testing of PUMPS and Valves, Revision 6
Miscellaneous
Engineering Permanent Modification Change Package No. 12179, Remote Racking Device -
4.16 kV 1E Switchgear NB001 and NB002, Revision 1
Temporary Modification Order 07-010-RP for 7300 System Cabinets 8 & 9, RP 044
Inservice Testing program Third 10-Year Interval, Containment Spray Pump Full Flow Testing
Line, Revision 5
-7- Attachment
WCOP-02, Revision 14, IST Program Plan
Section 1R19: Postmaintenance Testing
Procedures
AP 20E-001, Industry Operating Experience Program, Revision 9
ET 07-0054, 69 kV Transmission Line from Wolf Creek
MPE NE-002, Governor Adjustments For Emergency Diesel Generator NE02, Revision 8
MPE NE-003, Governor Adjustments For Emergency Diesel Generator NE01, Revision 7
MPM M018Q-01, Standby Diesel Generator Inspection, Revision 12
STN FP-211, "Diesel Driven Fire Pump 1FP01PB Monthly Operation and Fuel Level Check,"
Revision 15
STS KJ-015B, Manual/Auto Fast Start, Sync & Loading of EDG NE02, Revision 25A
STS KJ-015A, Manual/Auto Fast Start, Sync & Loading of EDG NE01, Revision 24
STS IC-615A, Slave Relay Test K615 Train A Safety Injection, Revision 20
STS BG-100B, Centrifugal Charging System B Train Inservice Pump Test, Revision 34
STS EJ-100B, RHR System Inservice Pump B Test, Revision 31
SYS KJ-123, Post Maintenance Run of Emergency Diesel Generator A, Revision 38
SYS KJ-200, Inoperable Emergency Diesel, Revision 13
Work Orders
07-299955-000 07-063761 07-301016-000 07-300862-001
07-300862-002 07-300768-001 07-301379-001 06-286736-001
06-286765-001 06-286737-001 07-298218-001
Condition Reports
2007-004117 2007-000279 2007-004190 2007-004117
2007-004190 2007-004471
-8- Attachment
Miscellaneous
Performance Improvement Request 2004-1160
Performance Improvement Request 2007-3829
Calculation XX-E-014, Analysis For NB Buses as Powered from Remote Generation,
Revision 0
Calculation XX-E-014 Attachment 9 OTI Sharpe Generation Station - Development & Testing
of ETAP User-Defined Dynamic Models (UDM), Revision 0
TMP 07-025, EJ FCV-611 Retest, Revision 0
TMP 07-014, BN HV-8812B Retest, Revision 0B
Section 1R22: Surveillance Testing
Procedures
STS AB-201D :Atmospheric Relief Valve Inservice Valve Test, Revision 20
STS EJ-100B, RHR System Inservice Pump B Test, Revision 31
STS GG-001A, Exhaust Filtration System Train A, 10-Hour Operability Test, Revision 19B
STS KJ-011A, DG NE01 24-Hour Run, Revision 19
ZL-005A, A EDG Operating Log, Revision 1A-Calculation sheet M-JE-321, Revision 2
Work Orders
07-296486-000 05-279238-000 05-279238-001 05-279238-002
05-279238-003 05-79238-004 36022
Section 1EP6: Drill Evaluation
Procedures
AP 06-002, Wolf Creek Nuclear Generating Station Emergency Plan, Revision 8
APF 21-001-02, Control Room Turnover Checklist
EPF 06-007-01, Wolf Creek Generating Station Emergency Notification, Revision 9
EPP 06-005, Emergency Classification, Revision 3
EPP 06-007, Emergency Notifications, Revision 12
EPP 06-011, Emergency Team Formation and Control, Revision 5
OFN NB-0034, Loss of All AC Power Shutdown Conditions, Revision 19
OFN NB-0030, Loss of AC Emergency Bus NB01 (NB02), Revision 10
Miscellaneous
Scenarios and Drill Evaluations, for drill conducted: January 31, 2008
Lesson LR 5004005 007, Loss of All AC While Shutdown, Revision 7
-9- Attachment
Section 2OS1: Access Controls to Radiologically Significant Areas and Section 2OS2:
ALARA Planning and Controls
Corrective Action Documents
2007-003381 2007-003904 2007-003932 2007-003934
2007-004065 2007-004139 2007-004183 2008-000104
2008-000112 2008-000883 2008-000980 2008-001181
2008-001077 2008-001304 2008-001336 2008-001346
2008-001349
Audit and Self-Assessment
QA Observation 14175; Radiological Controls, Radiological Postings
Radiation Work Permits
2008-0008 2008-3021 2008-1101
Procedures
RPP 01-105, Health Physics Organization, Responsibilities, and Code of Conduct, Revision 11
RPP 02-205, Radiological Survey Frequency Requirements, Revision 11
RPP 02-210, Radiation Survey Methods, Revision 29
RPP 02-215, Posting of Radiological Controlled Areas, Revision 23
RPP 02-405, RCA Access Control, Revision 14
AP 25A-001, Radiation Protection Manual, Revision 13
AP 25A-200, Access to Locked High or Very High Radiation Areas, Revision 15
AP 25B-200, Radiography Guidelines, Revision 11
Section 4OA1: Performance Indicator Verification
Procedures
AP 26A-007, NRC Performance Indicators, Revision 5
NEI 99-02, Regulatory Assessment Indicator Guideline, Revision 4
AP 26A-007, NRC Performance Indicators, Revision 5
AP 06-002-01, Emergency Action Levels, Revision 0
AP 06-002-01, Emergency Action Levels, Revision 10
AP 21-001, Conduct of Operations, Revision 37
Section 4OA2: Problem Identification and Resolution
Condition Reports
2006-000686 2006-001095 2006-001836 2006-001644
2006-002159 2006-002446 2007-002907 2007-003732
2007-002437 2007-003867 2006-002466 2006-002469
- 10 - Attachment
2006-003244 2006-004212 2007-000510 2007-001118
2007-04362 2007-002963 2007-002907 2006-001663
2007-001457 2007-001681 2007-002164 2007-002184
2007-002924 2007-004212 2007-001847 2007-002120
2007-003124 2007-004161 2007-004164 2007-004165
2007-004167 2007-004168 2007-004169 2007-004171
2007-004172 2007-004173 2007-004174 2007-004176
2007-004177 2007-004178 2007-004179 2007-004180
2007-004183 2007-004185 2007-004187 2007-004196
2007-004219 2007-002670 2006-002659 2008-001349
Work Orders
07-298655-000 06-288862-000 06-289411-000 07-293540-000
05-276746-001 05-276746-000 07-297825-000 01-227941-000
00-221564-000 07-293028-001 05-269169-000
Miscellaneous
Human Performance Initiative Status Report, December 2007
Operations Department Performance Indicators, March 2008
Wolf Creek Generating Station Performance Assessment Report, July through September 2007
Engineering Disposition 012487, Improvements on Intake manifold mounting and o-ring
capturing, Revision 0
Engineering Screening 012487, Improvements on Intake manifold mounting and o-ring
capturing, Revision 0
Work Request R 07-064173
Areva NP, GRW 06-044, October 6, 2006
ASCO Important Safety Notice, September 18, 2006
Performance Improvement Request 2001-0191
- 11 - Attachment
Section 4OA3: Event Followup
Calculations
AN 94-041, WCGS IPEEE Project IPEEE Fire Initiation Frequencies, Revision 0
AN 95-029, WCGS IPEEE Project Control Room Fire Analysis, Revision 1
AN 98-023, WCGS Fire Risk Evaluation Re-analysis, Revision 0
E-1F9900, Post-Fire Safe Shutdown Manual Actions, Revision 2
E-1F9905, Fire Hazard Analysis, Revision 0
E-1F9910, Post-Fire Safe Shutdown Fire Area Analysis, Revision 2
Condition Report
2006-00551
Drawings
E-1F3301, Fire Detection/Protection System Control Bldg, Diesel Gen Bldg, & Comm Corr,
-EL 2000'-0" & EL 2016'-0", Revision 4
E-1R3412, Exposed Conduit Control Building Area-1 El 2016'0", Revision 8
E-13KJ03A, Schematic Diagram Diesel Gen KKJ01B Engine Control (Start/Stop Circuit),
Revision 12
WIP-E-13CK13-004-A-1, Schematic Diagram Class 1E Electrical Equipment A/C Unit,
Revision 0
WIP-E-13KJ03A-012-A-1, Schematic Diagram Diesel Gen KKJ01B Engine Control (Start/Stop
Circuit), Revision 00
Problem Improvement Requests
2005-03033 2005-03209 2005-03314 2005-03333 2005-03364
Procedures
OFN KC-016, Fire Response, Revision 15
OFN RP-014, Hot Standby to Cold Shutdown from Outside the Control Room, Revision 9
OFN RP-017, Control Room Evacuation, Revisions 21, 22, 23, 24, & 25
Drawings
10466 A 1802, Architectural Fire Delineation Floor Plan EL. 2000' 0", Revision 13
E 1R1323A, Exposed Conduit, Auxiliary Building, Area 2 EL. 2000' 0", Revision 10
E 1R1323B(Q), Exposed Conduit, Auxiliary Building, Area 2 EL. 2000' 0", Revision 5
E 1R1323D, Exposed Conduit, Auxiliary Building, Area 2 EL. 2000' 0", Revision 6
E 1R1343A(Q), Exposed Conduit, Auxiliary Building, Area 4 EL. 2000' 0", Revision 1
E 1R1343B, Exposed Conduit, Auxiliary Building, Area 4 EL. 2000' 0", Revision 10
E 1R1343C, Exposed Conduit Auxiliary, Building, Area 4 EL. 2000' 0", Revision 12
E 1R1911, Raceway Sections & Details, Auxiliary Building, Revision 9
- 12 - Attachment
Miscellaneous
IEEE Standard 383-1974, IEEE Standard for Type Test of Class IE Electric Cables, Field
Splices, and Connections for Nuclear Power Generating Stations
Information Notice 2005-14, Fire Protection Findings on Loss of Seal Cooling to Westinghouse
Reactor Coolant Pumps, dated June 5, 2005
NUREG/CR-4527, An Experimental Investigation of Internally Ignited Fires in Nuclear Power
Plant Control Cabinets: Part 1: Cabinet Effects Tests, April 1987
NUREG/CR-4527, An Experimental Investigation of Internally Ignited Fires in Nuclear Power
Plant Control Cabinets: Part II: Room Effects Tests, November 1988
NUREG/CR-6850, EPRI/NRC-RES Fire PRA Methodology for Nuclear Power Facilities,
Volume 2: Detailed Methodology, September 2005
Specification 10466-J-200(Q), Technical Specification for Main Control Panels for the
Standardized Nuclear Unit Power Plant System, dated September 1979
Technical Bulletin TB-04-22, Reactor Coolant Pump Seal Performance - Appendix R
Compliance and Loss of All Seal Cooling, Revision 1
Updated Final Safety Analysis Report Section 9.2.1.2, Essential Service Water System
WCAP-16141, RCP Seal Leakage PRA Model Implementation Guidelines for Westinghouse
PWRS, August 2003
WCAP-16396-NP, Reactor Coolant Pump Seal Performance for Appendix R Assessments,
dated January 2005
Work Order 06-286793-000 Pre-outage Inspection of Rod Cluster Control Change Tool
Licensee Event Report 2005-005-00
Performance Improvement Request 2005-2757
Section 4OA5: Other Activities
AN 94-041, WCGS IPEEE Project IPEEE Fire Initiation Frequencies, Revision 0
AN 95-029, WCGS IPEEE Project Control Room Fire Analysis, Revision 1
AN 98-023, WCGS Fire Risk Evaluation Re-analysis, Revision 0
E-1F9900, Post-Fire Safe Shutdown Manual Actions, Revision 2
E-1F9905, Fire Hazard Analysis, Revision 0
E-1F9910, Post-Fire Safe Shutdown Fire Area Analysis, Revision 2
- 13 - Attachment
Drawings
5775-2, COMSIP Customline Corp Console (RL001 & RL002) Front Arrangement, Revision 19
5775-2, Main Control Console - RL001 & RL002 Plan, Rear, & Side, Elevation Plus Notes,
Revision 15, Sheet 2
5775-2, Main Control Console - RL001 & RL002 Sections Showing Equipment Clearance,
Revision 8, Sheet 4
5775-4, Operator Console RL005 & RL006 Front Arrangement, Revision 0
5775-4, Main Control Console - RL005 & RL006 Plan, Rear, & Side, Elevation Plus Notes,
Revision 15, Sheet 2
5775-4, Main Control Console - RL005 & RL006 Sections Showing Equipment Clearance,
Revision 6, Sheet 4
5775-7, COMSIP Customline Corp Main Control Board RL017 & RL018 Front Arrangement,
Revision 17
5775-7, Main Control Board - RL017 & RL018 Plan, Rear, & Side, Elevation & Notes,
Revision 14, Sheet 2
5775-8, Main Control Board RL019 & RL020 Front Arrangement," Revision 15, Sheet 3
5775-8, Main Control Board - RL019 & RL020 Plan, Rear, & Side, Elevation Plus Notes,
Revision 15, Sheet 2
E-13BG13, Schematic Diagram Boric Acid Filter to Charging Pump Valve, Revision 2
E-13EF07, Schematic Diagram ESW to Containment Air Coolers Isolation Valves, Revision 2
J-14001, Control Room Equipment Arrangement, Revision 6
WIP-E-13BG13-002-A-1, Schematic Diagram Boric Acid Filter to Charging Pump Valve,
Revision 0
WIP-E-13EF07A-000-A-1, Schematic Diagram ESW to Containment Air Coolers Isolation
Valves, Revision 0
Piping and Instrumentation Diagrams
M-12BB01, Reactor Coolant System, Revision 11
M-12BG01, Chemical and Volume Control System, Revision 14
M-12BG02, Chemical and Volume Control System, Revision 15
M-12BG03, Chemical and Volume Control System, Revision 37
M-12BG04, Chemical and Volume Control System, Revision 07
M-12BG05, Chemical and Volume Control System, Revision 13
M-12EF01, Essential Service Water System, Revision 20
M-12EF02, Essential Service Water System, Revision 23
- 14 - Attachment
M-12EG01, Component Cooling Water System, Revision 15
M-12EG02, Component Cooling Water System, Revision 18
M-12EG03, Component Cooling Water System, Revision 09
M-K2EF01, Essential Service Water System, Revision 49
M-K2EF03, Essential Service Water System, Revision 08
Problem Improvement Requests
2005-03033 2005-03209 2005-03314 2005-03333 2005-03364
Procedures
OFN-RP-017, Control Room Evacuation, Revisions 20-25
Miscellaneous
Diagrams of Control Panels RL001, -RL002, -RL005, -RL006, -RL0015, -RL016, -RL017,
-RL018, -RL019, and -RL020
Licensed Operator Lesson Plans related to auxiliary feedwater, chemical and volume control
system, component cooling water, and essential service water systems.
NUREG/CR-4527, An Experimental Investigation of Internally Ignited Fires in Nuclear Power
Plant Control Cabinets: Part 1: Cabinet Effects Tests, April 1987
NUREG/CR-4527, An Experimental Investigation of Internally Ignited Fires in Nuclear Power
Plant Control Cabinets: Part II: Room Effects Tests, November 1988
NUREG/CR-6850, EPRI/NRC-RES Fire PRA Methodology for Nuclear Power Facilities,
Volume 2: Detailed Methodology, September 2005
Specification 10466-J-200(Q), Technical Specification for Main Control Panels for the
Standardized Nuclear Unit Power Plant System, dated September 1979
Wolf Creek Generating Electric Station Individual Plant Examination of External Events (IPEEE)
Condition Reports
2007-001897 2007-002599
Work Orders
01-227795-000 07-296378-000 07-296378-001
Work Requests
07-061699 07-063138
Condition Reports
2007-003310 2007-002599 2007-001897
- 15 - Attachment
LIST OF ACRONYMS
ALARA as low as is reasonably achievable
ASME American Society of Mechanical Engineers
CCP centrifugal charging pump
CFR Code of Federal Regulations
CR condition report
ECCS emergency core cooling system
EDG emergency diesel generator
ESW essential service water
FIFCR fire ignition frequency for the control room
I&C instrumentation and control
IEEE Institute of Electrical and Electronics Engineers
IMC inspection manual chapter
IPEEE individual plant examination of external events
LER licensee event report
MCM million circular mils
NCV noncited violation
NEI Nuclear Energy Institute
NOED Notice of Enforcement Discretion
NPCRE control room evacuation
NRC Nuclear Regulatory Commission
PIR performance improvement request
SSC structure, system, and component
SER Safety Evaluation Report
TDAFW turbine-driven auxiliary feedwater
URI unresolved item
USAR Updated Safety Analysis Report
- 16 - Attachment