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{{#Wiki_filter:November 10, 2009 | {{#Wiki_filter:UNITED STATES | ||
NUC LE AR RE G UL AT O RY C O M M I S S I O N | |||
Rick A. Muench, President and | R E GI ON I V | ||
612 EAST LAMAR BLVD , SU I TE 400 | |||
AR LI N GTON , TEXAS 76011-4125 | |||
inspection at your Wolf Creek Generating Station. | November 10, 2009 | ||
The inspections examined activities conducted under your license as they relate to safety and compliance with the | Rick A. Muench, President and | ||
Chief Executive Officer | |||
This report documents eight NRC identified findings of very low safety significance (Green). | Wolf Creek Nuclear Operating Corporation | ||
of the noncited violations, you should provide a response within 30 days of the date of this inspection report, with the basis for your denial, to the U.S. Nuclear Regulatory Commission, ATTN: | P.O. Box 411 | ||
Burlington, KS 66839 | |||
SUBJECT: WOLF CREEK GENERATING STATION - NRC INTEGRATED INSPECTION | |||
REPORT 05000482/2009004 | |||
Dear Mr. Muench: | |||
On September 30, 2009, the U.S. Nuclear Regulatory Commission (NRC) completed an | |||
inspection at your Wolf Creek Generating Station. The enclosed integrated inspection report | |||
documents the inspection findings, which were discussed on October 14, 2009, with | |||
Mr. Matt Sunseri, Vice President of Operations and Plant Manager, and other members of your | |||
staff. | |||
The inspections examined activities conducted under your license as they relate to safety and | |||
compliance with the Commissions rules and regulations and with the conditions of your license. | |||
The inspectors reviewed selected procedures and records, observed activities, and interviewed | |||
personnel. | |||
This report documents eight NRC identified findings of very low safety significance (Green). All | |||
of these findings were determined to involve violations of NRC requirements. However, | |||
because of the very low safety significance and because they are entered into your corrective | |||
action program, the NRC is treating these findings as noncited violations, consistent with | |||
Section VI.A.1 of the NRC Enforcement Policy. If you contest the violations or the significance | |||
of the noncited violations, you should provide a response within 30 days of the date of this | |||
inspection report, with the basis for your denial, to the U.S. Nuclear Regulatory Commission, | |||
ATTN: Document Control Desk, Washington, D.C. 20555-0001, with copies to the Regional | |||
Administrator, U.S. Nuclear Regulatory Commission, Region IV, 612 E. Lamar Blvd, Suite 400, | |||
Arlington, Texas, 76011-4125; the Director, Office of Enforcement, U.S. Nuclear Regulatory | |||
Commission, Washington, D.C. 20555-0001; and the NRC Resident Inspector at the Wolf Creek | |||
Generating Station. | |||
In accordance with 10 CFR 2.390 of the NRC's "Rules of Practice," a copy of this letter, and its | |||
enclosure, will be available electronically for public inspection in the NRC Public Document | |||
Room or from the Publicly Available Records component of NRCs document system (ADAMS). | |||
Wolf Creek Nuclear Operating Corporation - 2 - | |||
ADAMS is accessible from the NRC Web site at http://www.nrc.gov/reading-rm/adams.html (the | |||
Public Electronic Reading Room). | |||
Wolf Creek Nuclear Operating Corporation - 2 - | Sincerely, | ||
/RA/ | |||
Geoffrey B. Miller, Chief | |||
Division of Reactor Projects | Project Branch B | ||
Division of Reactor Projects | |||
Docket No. 50-482 | |||
License No. NPF-42 | |||
Enclosure | |||
Inspection Report 05000482/2009004 | |||
w/Attachment: Supplemental Information | |||
cc w/Enclosure: | |||
Vice President Operations/Plant Manager | |||
Wolf Creek Nuclear Operating Corporation | |||
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 Corporation | |||
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 66612-1590 | |||
Attorney General | |||
120 S.W. 10th Avenue, 2nd Floor | |||
Topeka, KS 66612-1597 | |||
Wolf Creek Nuclear Operating Corporation - 3 - | |||
County Clerk | |||
Coffey County Courthouse | |||
110 South 6th Street | |||
Burlington, KS 66839 | |||
Chief, Radiation and Asbestos | |||
Control Section | |||
Bureau of Air and Radiation | |||
Kansas Department of Health and | |||
Environment | |||
1000 SW Jackson, Suite 310 | |||
Topeka, KS 66612-1366 | |||
Chief, Technological Hazards | |||
Branch | |||
FEMA, Region VII | |||
9221 Ward Parkway | |||
Suite 300 | |||
Kansas City, MO 64114-3372 | |||
Wolf Creek Nuclear Operating Corporation - 4 - | |||
Electronic Distribution by RIV: | |||
Regional Administrator (Elmo.Collins@nrc.gov) | |||
Deputy Regional Administrator (Chuck.Casto@nrc.gov) | |||
DRP Director (Dwight.Chamberlain@nrc.gov) | |||
DRP Deputy Director (Anton.Vegel@nrc.gov) | |||
DRS Director (Roy.Caniano@nrc.gov) | |||
DRS Deputy Director (Troy.Pruett@nrc.gov) | |||
Senior Resident Inspector (Chris.Long@nrc.gov) | |||
Resident Inspector (Charles.Peabody@nrc.gov) | |||
Site Secretary (Shirley.Allen@nrc.gov) | |||
Branch Chief, DRP/B (Geoffrey.Miller@nrc.gov) | |||
Senior Project Engineer, DRP/B (Rick.Deese@nrc.gov) | |||
Public Affairs Officer (Victor.Dricks@nrc.gov) | |||
Technical Support Branch (Chuck.Paulk@nrc.gov) | |||
RITS Coordinator (Marisa.Herrera@nrc.gov) | |||
Regional Counsel (Karla.Fuller@nrc.gov) | |||
Congressional Affairs Officer (Jenny.Weil@nrc.gov) | |||
OEMail Resource | |||
Regional State Liaison Officer (Bill.Maier@nrc.gov) | |||
NSIR/DRP/EP (Robert.Kahler@nrc.gov) | |||
DRS STA (Dale.Powers@nrc.gov) | |||
OEDO RIV Coordinator (Leigh.Trocine@nrc.gov) | |||
ROPreports | |||
File located: R:\_REACTORS\_WC\2009\WC 2009-004RP-CML.doc ADAMS ML093140803 | |||
SUNSI Rev Compl. 7Yes No ADAMS 7Yes No Reviewer Initials GM | |||
Publicly Avail 7Yes No Sensitive Yes 7 No Sens. Type Initials GM | |||
RI:DRP/ SRI:DRP/ SPE:DRP/ C:DRS/EB1 C:DRS/EB2 | |||
CAPeabody CMLong PJayroe RLKellar NFOKeefe | |||
/RA - E/ /RA - E/ /RA/ /RA WSifre for/ /RA/ | |||
10/20/2009 10/20/2009 11/10/2009 11/05/2009 11/05/2009 | |||
C:DRS/OB C:DRS/PSB1 C:DRS/PSB2 C:DRP/ | |||
SGarchow MPShannon GEWerner GBMiller | |||
/RA/ /RA Johnson for/ /RA/ /RA/ | |||
11/05/2009 11/05/2009 11/06/2009 11/10/2009 | |||
OFFICIAL RECORD COPY T=Telephone E=E-mail F=Fax | |||
U.S. NUCLEAR REGULATORY COMMISSION | |||
REGION IV | |||
Docket: 50-482 | |||
License: NPF-42 | |||
Report: 05000482/2009004 | |||
Licensee: Wolf Creek Operating Corporation | |||
Facility: Wolf Creek Generating Station | |||
Location: 1550 Oxen Lane SE | |||
Burlington, Kansas | |||
Dates: July 1 through September 30, 2009 | |||
Inspectors: C. M. Long, Senior Resident Inspector | |||
C. A. Peabody, Resident Inspector | |||
P. A. Jayroe, Project Engineer, Project Branch B | |||
G. W. Apger, Operations Engineer | |||
S. M. Alferink, Reactor Inspector, Engineering Branch 2 | |||
J. P. Adams, Reactor Inspector, Engineering Branch 1, DRS | |||
C. M. Ryan, Reactor Inspector, Engineering Branch 1, DRS | |||
G. P. Tutak, Reactor Inspector, Engineering Branch 2, DRS | |||
Approved By: G. B. Miller, Chief, Project Branch B | |||
Division of Reactor Projects | |||
-1- Enclosure | |||
SUMMARY OF FINDINGS | |||
IR 05000482/2009004, 7/1/2009 - 9/30/2009; Wolf Creek Generating Station, Integrated | |||
Resident and Regional Report; Operability Evaluations; Post Maintenance Testing; Plant | |||
Modifications; Maintenance Risk Assessments and Emergent Work Control; Fire Protection; | |||
Event Followup. | |||
The report covered a 3-month period of inspection by resident inspectors and an announced | |||
baseline inspections by regional based inspectors. Seven Green and one Severity Level IV | |||
noncited violations of significance were identified. The significance of most findings is indicated | |||
by their color (Green, White, Yellow, or Red) using Inspection Manual Chapter 0609, | |||
Significance Determination Process. Findings for which the significance determination | |||
process does not apply may be Green or be assigned a severity level after NRC management | |||
review. The NRC's program for overseeing the safe operation of commercial nuclear power | |||
reactors is described in NUREG 1649, Reactor Oversight Process, Revision 4, dated | |||
December 2006. | |||
A. NRC-Identified Findings and Self-Revealing Findings | |||
Cornerstone: Mitigating Systems | |||
* Green. On June 30, 2009, the inspectors identified a noncited violation of Technical | |||
Specification 3.8.1 for failure to perform an adequate common cause evaluation within | |||
24 hours to demonstrate no common cause failure mechanism existed between the | |||
emergency diesel generators after a through-wall leak was discovered on the essential | |||
service water piping. Wolf Creek did not start the opposite train emergency diesel | |||
generator and declared that the through-wall flaw was not a common cause failure | |||
without any evaluation or supporting statements. Nondestructive testing had not been | |||
started at this time. Subsequent evaluation of the flaw per American Society of | |||
Mechanical Engineers (ASME) Code Case N513.2 restored operability to the essential | |||
service water piping. The licensee entered this issue in their corrective action program | |||
as Condition Report 18347. | |||
The inspectors determined that the failure to demonstrate, per Technical | |||
Specifications 3.8.1 Required Actions B.3.1 or B.3.2, that no common cause failure | |||
existed for the emergency diesel generators 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 | |||
affected the cornerstone objective to ensure the availability, reliability, and capability of | |||
systems that respond to initiating events to prevent undesirable consequences. The | |||
inspectors evaluated the significance of this finding using Phase 1 of Inspection Manual | |||
Chapter 0609, Appendix A, "Significance Determination of Reactor Inspection Findings | |||
for At Power Situations," and determined that the finding was of very low safety | |||
significance (Green) because the issue was not a design or qualification deficiency | |||
confirmed to result in loss of operability or functionality, did not represent a loss of | |||
system safety function, an actual loss of safety function of a single train for greater than | |||
its technical specification allowed outage time, an actual loss of safety function of a | |||
nontechnical specification risk-significant equipment train, and did not screen as | |||
potentially risk significant due to a seismic, flooding, or severe weather initiating event. | |||
-2- Enclosure | |||
The cause of the finding has a problem identification and resolution crosscutting aspect | |||
in the area associated with the corrective action program because Wolf Creek failed to | |||
thoroughly evaluate the failure mechanism such that the resolutions address the causes | |||
and extent of conditions, as necessary. Specifically Wolf Creek did not properly | |||
consider the possibility of common-cause pitting failures which could have impacted the | |||
essential service water piping Train A structural integrity thereby affecting its cooling | |||
loads, including the Emergency Diesel Generator A [P.1(c)] (Section 1R15). | |||
* Green. The inspectors identified a noncited violation of Technical Specification 3.8.1, | |||
Required Action B.4.2.2 on March 24, 2009 when the licensee performed elective | |||
maintenance on safety bus relays and removed equipment from service that was | |||
required by the technical specification | |||
a. Inspection Scope | a. Inspection Scope | ||
The inspectors reviewed the effectiveness of the licensees implementation of | |||
evaluations performed in accordance with 10 CFR 50.59, Changes, Tests, and | |||
Experiments, and changes, tests, experiments, or methodology changes that the | |||
licensee determined did not require 10 CFR 50.59 evaluations. The inspection | |||
procedure requires the review of 6 to 12 licensee evaluations required by 10 CFR 50.59, | |||
12 to 25 changes, tests, or experiments that were screened out by the licensee and 5 to | |||
15 permanent plant modifications. | |||
The inspectors reviewed 9 evaluations required by 10 CFR 50.59. These included: | |||
* 2006-001, Radiological Consequences of a Fuel Handling Accident, Revision 0 | |||
* 2008-0006, Wolf Creek Generating Station (WCGS) Simplified Head Assembly | |||
(SHA) Drop Analysis, Revision 0 | |||
* 2008-0008, Use of Dedicated Operator for SI Pump B Room cooler | |||
Replacement, Revision 0 | |||
* 2005-004, WCGS Rod Withdrawal at Power Event Safety Analysis, Revision 0 | |||
* 2008-001, Evaluations of Voids in the ECCS Suction Piping, Revision 0 | |||
* 2008-002, Evaluations of Voids in the ECCS Discharge Piping, Revision 0 | |||
* 2006-002, Power Operation, Revision 54 | |||
- 21 - Enclosure | |||
* 2008-0003, Use of Dedicated Operator for SI Pump A Room Cooler | |||
Replacement , Revision 0 | |||
* 2008-0004, MSFIS Controls Replacement , Revision 0 | |||
The inspectors reviewed 17 changes, tests, and experiments that were screened out by | |||
licensee personnel. These included: | |||
* CP 12731, RCP No. 1 Seal Housing Stud Preload Evaluation, Revision 0 | |||
* CP 12746, Torque of Piping Flanges Between EDG Heat Exchangers, Revision 1 | |||
* CP 12820, Containment Room Cooler SGN01D, Revision 1 | |||
* CP 12876, Main Steam Atmospheric Relief Valve Aux (Pilot) Plug and Main Plug | |||
Machining Dimensions, Revision 0 | |||
* CP 12979, Updating the RCS pressure and temperature limits, PORV lift setting | |||
for the LTOP system, and the PTLR, Revision 0 | |||
* CP 13089, EF-138-HBC-30 Essential Service Water Pipe Pit Encapsulation, | |||
Revision 1 | |||
* CP 11987, EKJ03A/B Replacement Heat Exchangers, Revision 6 | |||
* CP 12758, Coating Degradation and Isolated Pitting of Containment Incore | |||
Instrumentation Sump Layer, Revision 3 | |||
* CP 12240, Over Torque on Valve GTHZ0008, Revision 0 | |||
* CP 12273, Shrinkage Effect at the Pressurizer Spray Nozzle on TBB03 Due to | |||
Weld-Overlay, Revision 3 | |||
* CP 12489, SGK05A Tube Sheet and Channel Cover Degradation Evaluation, | |||
Revision 0 | |||
* CP 12341, Region 19 Fuel Assembly and Core Component Configuration | |||
Changes, Revision 0 | |||
* CP 12154, Relocate CVT Level transmitter BGLT0185, Revision 3 | |||
* CP 12175, PFSSD MOV Hot Short Mod: BGHV8111, BNLCV0112E, | |||
EMHV8803B, Revision 0 | |||
* CP 12639, 9 Volt Power Supply for SP067 & SP010 , Revision 0 | |||
* CP 12782, NE107187 DG NE01 Generator Differential relay, Revision 0 | |||
- 22 - Enclosure | |||
* TMP 08-022 , SI Accumulator C Boron Concentration Adjustment, Revision 0 | |||
The inspectors reviewed 7 permanent plant modifications. These included: | |||
* CP 11987, EKJ03A/B Replacement Heat Exchangers, Revision 6 | |||
* CP 11379, Replacement for Obsolete Rad Monitoring Transducer, Revision 2 | |||
* CP 13089, EF-138-HBC-30" Essential Service Water Pipe Pit Encapsulation, | |||
Revision 1 | |||
* CP 12673, Installation of Vents in the Bonnets of EJ8958A and EJ8958B, | |||
Revision 1 | |||
* CP 9488, Governor Replacement on Emergency Diesel Generators, Revision 7 | |||
* CP 11608, MSIV and MFIV Actuator Replacement Electrical Work, Revision 10 | |||
* CP 11897, Transformer XNB02 Tap Change, Revision 2 | |||
The inspectors verified that when changes, tests, or experiments were made, that | |||
evaluations were performed in accordance with 10 CFR 50.59 and that licensee | |||
personnel had appropriately concluded that the change, test or experiment can be | |||
accomplished without obtaining a license amendment. The inspectors also verified that | |||
safety issues related to the changes, tests, or experiments were resolved. The | |||
inspectors reviewed changes, tests, and experiments that licensee personnel | |||
determined did not require evaluations and verified that the | |||
stated that they had reviewed proprietary information during the inspection, and verified | |||
that all material had been returned to the licensee or destroyed. The licensee | |||
acknowledged the inspection results as presented. | |||
The inspector briefed Robert Evenson of the results of the annual licensed operator | |||
requalification program inspection on August 5, 2009. The licensee representative | |||
acknowledged the findings presented. The inspectors asked the licensee whether any | |||
materials examined during the inspection should be considered proprietary. No | |||
proprietary information was identified. | |||
On September 28, 2009, the inspectors conducted a telephonic exit meeting and | |||
presented the results of the staff review of fire protection program changes to | |||
Mr. J. Suter, Fire Protection Supervisor, and other members of the licensee staff. The | |||
licensee acknowledged the issues presented. The inspectors asked the licensee | |||
whether any of the material examined during the inspection should be considered | |||
proprietary. No proprietary information was identified. | |||
On October 14, 2009, the resident inspectors presented the inspection results of the | |||
resident inspections to Mr. Matt Sunseri, Vice President Oversight, and other members | |||
of the licensee's management staff. The licensee acknowledged the findings presented. | |||
The inspectors noted that while proprietary information was reviewed, none would be | |||
included in this report and that the materials were returned to the licensee. | |||
- 44 - Enclosure | |||
SUPPLEMENTAL INFORMATION | |||
KEY POINTS OF CONTACT | |||
Licensee Personnel | |||
R. A. Muench, President and Chief Executive Officer | |||
M. Sunseri, Vice President Operations and Plant Manager | |||
S. E. Hedges, Vice President Oversight | |||
G. J. Pendergrass, Manager Engineering | |||
T. East, Manager, Emergency Planning | |||
P. Bedgood, Superintendent, Chemistry/Radiation Protection | |||
LIST OF ITEMS OPENED AND CLOSED | |||
Opened and Closed | |||
05000482/2009004-01 NCV Failure to Log Foreign Material in Spent Fuel Pool After | |||
Extent of Condition Evaluation (Section 1R05) | |||
05000482/2009004-02 NCV Inability to Perform Manual Actions for Risk Assessment | |||
(Section 1R13) | |||
05000482/2009004-03 NCV Inadequate Evaluation of Emergency Diesel Generator | |||
for Common Cause Failure in the Supporting Essential | |||
Service Water System (Section 1R15.1) | |||
05000482/2009004-04 NCV Failure to Implement Engineered Safety Features | |||
Actuation System Technical Specifications Results in | |||
Missed Mode Change (Section 1R15.2) | |||
05000482/2009004-05 NCV Use of Nonsafety-Related Power to Ensure Operability of | |||
Safety-Related Boric Acid System (Section 1R17) | |||
05000482/2009004-06 NCV Performing Prohibited Elective Maintenance on Safety | |||
Bus NB02 Channel 4 During Emergency Diesel | |||
Generator Maintenance (Section 1R19) | |||
05000482/2009004-07 NCV Failure to Report Conditions that Could have Presented | |||
Fulfillment of a Safety Function (Section4OA3) | |||
05000482/2009004-08 NCV Changes to the Approved Fire Protection Program | |||
Without Prior Staff Approval (Section 4OA5.3) | |||
A-1 Attachment | |||
Closed | |||
05000482/2005008-02 AV Failure to Maintain Reactor Coolant System Subcooling | |||
During the Alternative Shutdown (Section 4OA5.4) | |||
05000482/2008010-03 URI Changes to the Approved Fire Protection Program May | |||
Not Meet NRC Acceptance Criteria (Section 4OA5.3) | |||
LIST OF DOCUMENTS REVIEWED | |||
Section 1RO1: Adverse Weather Protection | |||
DOCUMENTS | |||
NUMBER TITLE REVISION | |||
STS NB-005 Breaker Alignment Verification Revision 18 | |||
Section 1RO4: Equipment Alignment | |||
DOCUMENTS | |||
NUMBER TITLE REVISION | |||
CKL AL-120 Auxiliary Feedwater Normal Lineup 34 | |||
M-12AL01 Piping and Instrumentation Diagram - Auxiliary 10 | |||
Feedwater System | |||
M-12EF01 Piping and Instrumentation Diagram - Essential 21 | |||
Service Water System | |||
M-12EF02 Piping and Instrumentation Diagram - Essential 25 | |||
Service Water System | |||
M-12AB01 Piping and Instrumentation Diagram - Main Steam 11 | |||
System | |||
M-12AB02 Piping and Instrumentation Diagram - Main Steam 12 | |||
System | |||
A-2 Attachment | |||
Section 1RO4: Equipment Alignment | |||
DOCUMENTS | |||
NUMBER TITLE REVISION | |||
USAR 15.6-12/13, Steam Generator Tube Rupture 22 | |||
with Postulated Stuck-Open Atmospheric Relief Valve | |||
Control Room Logs dated September 16, 2009 at | |||
1:49 a.m. | |||
M-224A-00037 10 -900 Carbon Steel Flex Wedge Gate Valve with 6:1 G | |||
B.G. Actuator | |||
USAR Figure 9.3-8-03, Piping and Instrumentation 41 | |||
Diagram Chemical & Volume Control System | |||
Condition Reports | |||
00019813 00019821 00019825 | |||
Work Order | |||
09-3160637-000 | |||
Work Request | |||
09-072489 | |||
Section 1RO5: Fire Protection | |||
DOCUMENTS | |||
NUMBER TITLE REVISION | |||
AP 10-106 Fire Preplans 8 | |||
FPPM-009 Control Bldg El.2000 2 | |||
FPPM-014 Diesel Generator Rooms El.2000 1 | |||
A-3 Attachment | |||
Section 1R11: Licensed Operator Requalification Program | |||
DOCUMENTS | |||
NUMBER TITLE REVISION / | |||
DATE | |||
LR5004004 Shutdown LOCA 009 | |||
Section 1R13: Maintenance Risk Assessment and Emergent Work Controls | |||
DOCUMENTS | |||
NUMBER TITLE REVISION / | |||
DATE | |||
Week of 3/16/09 - Operational Risk Assessment | |||
Condition Reports | |||
00016735 00015318 2009-001338 | |||
Section 1R15: Operability Evaluations | |||
SYS SR-200 Moveable Incore Detector Operation 21 STS RE-012 QPTR Determination 10 STS RE-013C BEACON SinglePoint AFD Calibration 10 WO 09-318203-002 Engineering Disposition: | DOCUMENTS | ||
NUMBER TITLE REVISION / | |||
DATE | |||
STS SF002 Core Axial Flux difference 9 | |||
STS RE-009 Heat Flux Hot Channel Factor Measurement 14 | |||
SYS SR-200 Moveable Incore Detector Operation 21 | |||
STS RE-012 QPTR Determination 10 | |||
STS RE-013C BEACON SinglePoint AFD Calibration 10 | |||
WO 09-318203-002 Engineering Disposition: EF138HBC-30 has a thru June 30, 2009 | |||
wall leak | |||
WO 09-318203-009 Engineering Disposition: Minimum Wall Issues with July 16, 2009 | |||
Line EF138HBC-30 | |||
A-4 Attachment | |||
Section 1R15: Operability Evaluations | |||
DOCUMENTS | |||
NUMBER TITLE REVISION / | |||
DATE | |||
SWO 09-318982-001 Engineering Disposition: EF150HBC-18 Essential July 28, 2009 | |||
Service Water Pipe Pit Through Wall Leak | |||
WO 09-319429-001 Engineering Disposition: EF049HBC-8 Thru Wall Leak August 20, 2009 | |||
Evaluation | |||
SYS KJ-200 Inoperable Emergency Diesel 15 / June 30, | |||
2009 | |||
NRC GL 93-05 Line-Item Technical Specifications Improvements to September 27, | |||
Reduce Surveillance Requirements for Testing During 1993 | |||
Power Operation | |||
NPF-42 Amendment #101 August 9, 1996 | |||
NPF-42 Amendment #123 December 31, | |||
1999 | |||
STS IC-232 Channel Operational Test Nuclear Instrumentation 15 | |||
System Source Range N-32 Protection Set II | |||
Class IE Environmental Qualification Data Sheet for NI September 1990 | |||
31 and 32 Source Range Monitors | |||
Condition Reports | |||
00018217 00018347 00018611 00018945 00019276 | |||
00019282 00019307 | |||
Work Orders | |||
09-318203-001 (Ultrasonic Thickness Report) | |||
09-318268-000 (Ultrasonic Thickness Report) | |||
09-318269-000 (Ultrasonic Thickness Report) | |||
09-318270-000 (Ultrasonic Thickness Report) | |||
09-318271-000 (Ultrasonic Thickness Report) | |||
09-318272-000 (Ultrasonic Thickness Report) | |||
09-318982-000 (Ultrasonic Thickness Report) | |||
09-318982-003 (Ultrasonic Thickness Report) | |||
09-318982-013 (2 Ultrasonic Thickness Reports) | |||
A-5 Attachment | |||
09-318982-014 (3 Ultrasonic Thickness Reports) | |||
09-319473-000 (Ultrasonic Thickness Report) | |||
09-319473-001 (Ultrasonic Thickness Report) | |||
09-319473-002 (Ultrasonic Thickness Report) | |||
09-319473-003 (Ultrasonic Thickness Report) | |||
09-319473-004 (Ultrasonic Thickness Report) | |||
09-319473-006 (Ultrasonic Thickness Report) | |||
09-319473-007 (Ultrasonic Thickness Report) | |||
09-319473-008 (Ultrasonic Thickness Report) | |||
09-319473-009 (Ultrasonic Thickness Report) | |||
09-319473-010 (Ultrasonic Thickness Report) | |||
Section 1R17: Permanent Plant Modifications (71111.17A) | |||
Calculations | |||
Number Title Revision | |||
M-628-00131-W01 Control Logic Diagram MSIV PPS-700 0 | |||
M-630-0095-W01 Control Logic Diagram MFIV PPS-300 0 | |||
XX-E-013 Post-Fire Safe Shutdown Analysis 1 | |||
XX-E-016 XNB02 Tap Change Analysis 0 | |||
XX-E-006 AC System Analysis 5 | |||
Wolf Creek Generating Station Rod Withdrawal at Power | |||
AN-06-007 0 | |||
(RWAP) Event Safety Analysis | |||
AN-04-015 Radiological Consequences of a Fuel Handling Accident 1 | |||
Wolf Creek Generating Station (WCGS) Simplified Head | |||
0720517.01-C-001 0 | |||
Assembly (SHA) Drop Analysis | |||
Installation of Vent Lines on Check Valves EJ8958A, | |||
EJ-S-008 0 | |||
EJ8958B and EJ8958C | |||
Westinghouse Class I Nuclear Valves 6 and Larger | |||
XX-S-036 0 | |||
Swing Check Valves - EM5093 | |||
Condition Reports | |||
2006-000363 2006-000577 2006-001070 2006-001447 | |||
2006-001858 2006-001923 2006-002412 2006-003067 | |||
2006-003135 2006-003235 2006-003241 2007-000070 | |||
2007-000235 2007-000416 2007-001115 2007-002153 | |||
2007-002251 2007-002329 2007-002401 2007-002459 | |||
2007-002727 2007-003578 2007-003767 2007-003782 | |||
A-6 Attachment | |||
2007-004696 2008-000028 2008-000083 2008-000662 | |||
2008-000826 2008-001445 2008-001727 2008-002157 | |||
2008-004744 2008-005500 2008-005550 2008-005808 | |||
2009-000409 00014799 00016231 2007-001180 | |||
2006-000309 2006-000442 2006-001447 2007-001457 | |||
2006-001549 2006-003684 | |||
Drawings | |||
Number Title Revision | |||
WIP-E-15000- Electrical Cable, Termination, and Raceway List 5 | |||
065-R-1 | |||
E-13AB32 Miscellaneous Circuits 7 | |||
E-11025 Relay Settings Tabulation and Coordination Curves System 13 | |||
NE | |||
0405-0003-01 Intercooler Heat Exchanger Analysis Input Data 2 | |||
Miscellaneous | |||
Number Title Date/Revision | |||
USA 50.59 Resource Manual 3 | |||
NEI 96-07 Guidelines for 10 CFR 50.59 Implementation 1 | |||
J-200B-00001 Nutherm Qualification Report Eaton Cutler-Hammer 0 | |||
Contact Blocks With Separation Barriers | |||
PSA PCR 2006- Maintenance of the Wolf Creek PSA Model 0 | |||
0002 | |||
PSA 05-0002 WCGS PRA Initiating Event Notebook - 2002 Update 0 | |||
M-018B-00001 Instruction Manual for Governor Modification W03 | |||
Design Change Process Improvements Engineering | |||
N/A 0 | |||
Initiative Plan | |||
Design Change Process Improvement Initiative: Monthly | |||
N/A April 10, 2009 | |||
Progress Report | |||
A-7 Attachment | |||
Procedures | |||
Number Title Revision | |||
GEN 00-004 Power Operation 54 | |||
OFN RP-017 Control Room Evacuation 29 | |||
SYS EP-200 Safety Injection Accumulator Operations 30 | |||
AP 05-001 Change Package Planning and Implementation 7 | |||
AP 05-002 Dispositions and Change Packages 8 | |||
Design, Implementation & Configuration Control of | |||
AP 05-005 13 | |||
Modifications | |||
AP 05F-001 Design Verification 3 | |||
AP 26A-003 10 CFR 50.59 Reviews 10 | |||
Section 1R18: Plant Modifications | |||
NUMBER TITLE DATE | |||
09-005-XX-01 Temporary Modification Order February 19, 2009 | |||
09-008-SG00 Temporary Modification Order March 5, 2009 | |||
09-0019 Essential Required Reading: Responding to an March 12, 2009 | |||
Earthquake with Inoperable Seismic Instrumentation | |||
Change Package No. 011613 | |||
Condition Reports | |||
2009-001278 2009-001194 | |||
Work Requests | Work Requests | ||
09-072504 09-072505 09-072506 09-072507 09-072508 | |||
A-8 Attachment | |||
Section 1R19: Postmaintenance Testing | |||
DOCUMENTS | |||
NUMBER TITLE REVISION / | |||
DATE | |||
STS KJ-015B Manual/Auto Fast Start, Sync & Loading of EDG NE02 27A / | |||
August 17, | |||
2009 | |||
STS AL-103 Turbine Driven Auxiliary Feedwater Pump Inservice Pump 44 / | |||
Test September | |||
9, 2009 | |||
NP-1490 4-900 ANSI Trip Throttle Valve A | |||
103171D Trip Throttle Valve Electrical Schematic Sheet 1 June 5, | |||
1977 | |||
103171D Trip Throttle Valve Electrical Schematic Sheet 2 November | |||
17, 1980 | |||
Work Orders | |||
09-316773-000 09-316773-001 | |||
Section 1R20: Refueling and Other Outage Activities | |||
DOCUMENTS | |||
NUMBER TITLE REVISION | |||
Feedwater Isolation Logic Drawing | |||
Table 7.3.15 USAR - NSSS Interlocks for Engineered Safety Feature 13 | |||
Actuation System | |||
SYS SB-122 Enabling/Disabling P-4/LO Tavg Fwis 1 | |||
Table 7.5-1 Engineered Safety Features - Displays 21 | |||
12.2-7 Westinghouse Technology Systems Manual Reactor 0100 | |||
Protection system - Reactor Trip Signals | |||
A-9 Attachment | |||
Condition Reports | SYS SB-122 Enable/Disabling P-4/LO Tavg FWIS 1 | ||
7.2-31 USAR - Testing of Reactor Trip Breakers 11 | |||
Work Orders | |||
09-314863-002 09-319404-000 | |||
Performance Improvement Request | |||
2001-0041 | |||
Condition Reports | |||
00019318 00019318 | |||
Section 1R22: Surveillance Testing | |||
PROCEDURES | |||
NUMBER TITLE REVISION / | |||
DATE | |||
STS CR-004 Shift Log for Additional Monitoring 0 | |||
STS EF-100B ESW System Inservice Pump B & ESW B Discharge Check 32 / August | |||
Valve Test 13, 2009 | |||
STS IC-208B 4 kV Loss of Voltage and Degraded Voltage TADOT NB02 2A /July 14, | |||
Bus - Separation Group 4 2009 | |||
STS RE-006 End of Life Core Moderator Temperature Coefficient 18 / August | |||
Measurement 28, 2009 | |||
Work Order | |||
09-315436-000 | |||
Condition Reports | |||
00019069 00019000 | |||
A-10 Attachment | |||
Section 4OA1: Performance Indicator Verification | |||
PROCEDURES | |||
NUMBER TITLE REVISION | |||
STS IC-203 Channel Operational Test 7300 Process Instrumentation 22B | |||
Protection Set III (Blue) | |||
INC C-001 7300 Signal Comparator Card (NAL 1) 6 | |||
MISCELLANEOUS DOCUMENTS | |||
NUMBER TITLE REVISION | |||
NEI 99-02 Mitigating Systems Cornerstone 5 | |||
LER 2005-004-00 Failure of Auxiliary Building Ventilation Dampers to | |||
Close on Safety Injection Signal | |||
OPR01 Operability/Reportability Detail Report | |||
LER 2008-007-00 Two Residual Heat Removal Trains Inoperable in | |||
Mode 3 due to Check Valve Leakage | |||
LER 2008-004-00 Loss of Power Event When the Reactor was De- | |||
fueled | |||
LER 2008-008-01/02 Potential for Residual Heat Removal Trains to Be | |||
Inoperable During Mode Change | |||
LER 2008-009-00 Inadequate Compensatory Actions for a Fire Area | |||
LER 2008-001-00 Containment Cooler Inoperability (Callaway Plant | |||
Unit 1) | |||
AIF 16C-001-02 Maintenance Walkdown Form (Technician) 0 | |||
10466-M-761-2076- Interconnecting Wiring Diagram Cabinet 03 SNUPPS | |||
W05 Nuclear Power Plant Controls | |||
2000801894 Adverse Condition- Ameren | |||
A-11 Attachment | |||
MISCELLANEOUS DOCUMENTS | |||
NUMBER TITLE REVISION | |||
NCV Failure to Ensure the Suitability of the Design of the | |||
05000483/2008003-01 Containment Air Cooler control Circuitry | |||
Appendix D, 10 CFR 50.72 Including Statement of | |||
Considerations | |||
Event Notification Report of June 23, 2008 | |||
Condition Reports | |||
2009-00017786 2009-00017846 2009-00017851 2009-00019914 2009-00019371 | |||
2009-001326 2009-00017776 2009-0001261 2009-001326 2009-001004 | |||
2008-001307 2009-00018156 200-00018156 2008-000470 2008-001673 | |||
Work Orders | |||
09-314726-000 09-317948-000 09-306203-000 | |||
Corrective Action Plan | |||
4160 1970 3944 3943 | |||
Reportability Evaluation Request | |||
2008-011 2009-012 | |||
Section 4OA2: Identification and Resolution of Problems | |||
PROCEDURES | |||
NUMBER TITLE REVISION / | |||
DATE | |||
TMP 09-014 CCW Flow Balance for Troubleshooting Thermal Barrier 0 / July 15, | |||
Closure 2009 | |||
SYS EG-201 Transferring Supply of CCW Service Loop and CCW Train 36 / July 15, | |||
Shutdown 2009 | |||
A-12 Attachment | |||
Applicability Determination for TMP 09-014 July 14, | |||
2009 | |||
50.59 Screen for TMP 09-014 July 14, | |||
2009 | |||
USAR USAR Section 5.4.1.2.2 0 | |||
Work Order | |||
09-316483-000 | |||
Corrective Action | |||
00018793 | |||
Section 4OA5: Other Activities | |||
PROCEDURES | |||
NUMBER TITLE DATE | |||
SLNRC 84-109 Letter to NRC 08/23/1984 | |||
USAR CR Updated Final Safety Analysis Report Change Request 09/25/2008 | |||
2008-009 | |||
Evaluation of Proposed Change for USAR CR 2008-009 09/25/2008 | |||
Performance Improvement Request | |||
2008-004869 | 2008-004869 | ||
A-13 Attachment | |||
}} | }} |
Latest revision as of 01:36, 14 November 2019
ML093140803 | |
Person / Time | |
---|---|
Site: | Wolf Creek ![]() |
Issue date: | 11/10/2009 |
From: | Geoffrey Miller Division of Nuclear Materials Safety IV |
To: | Muench R Wolf Creek |
References | |
IR-09-004 | |
Download: ML093140803 (61) | |
See also: IR 05000482/2009004
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
November 10, 2009
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 05000482/2009004
Dear Mr. Muench:
On September 30, 2009, the U.S. Nuclear Regulatory Commission (NRC) completed an
inspection at your Wolf Creek Generating Station. The enclosed integrated inspection report
documents the inspection findings, which were discussed on October 14, 2009, with
Mr. Matt Sunseri, Vice President of Operations and Plant Manager, and other members of your
staff.
The inspections examined activities conducted under your license as they relate to safety and
compliance with the Commissions rules and regulations and with the conditions of your license.
The inspectors reviewed selected procedures and records, observed activities, and interviewed
personnel.
This report documents eight NRC identified findings of very low safety significance (Green). All
of these findings were determined to involve violations of NRC requirements. However,
because of the very low safety significance and because they are entered into your corrective
action program, the NRC is treating these findings as noncited violations, consistent with
Section VI.A.1 of the NRC Enforcement Policy. If you contest the violations or the significance
of the noncited violations, you should provide a response within 30 days of the date of this
inspection report, with the basis for your denial, to the U.S. Nuclear Regulatory Commission,
ATTN: Document Control Desk, Washington, D.C. 20555-0001, with copies to the Regional
Administrator, U.S. Nuclear Regulatory Commission, Region IV, 612 E. Lamar Blvd, Suite 400,
Arlington, Texas, 76011-4125; the Director, Office of Enforcement, U.S. Nuclear Regulatory
Commission, Washington, D.C. 20555-0001; and the NRC Resident Inspector at the Wolf Creek
Generating Station.
In accordance with 10 CFR 2.390 of the NRC's "Rules of Practice," a copy of this letter, and its
enclosure, will be available electronically for public inspection in the NRC Public Document
Room or from the Publicly Available Records component of NRCs document system (ADAMS).
Wolf Creek Nuclear Operating Corporation - 2 -
ADAMS is accessible from the NRC Web site at http://www.nrc.gov/reading-rm/adams.html (the
Public Electronic Reading Room).
Sincerely,
/RA/
Geoffrey B. Miller, Chief
Project Branch B
Division of Reactor Projects
Docket No. 50-482
License No. NPF-42
Enclosure
Inspection Report 05000482/2009004
w/Attachment: Supplemental Information
cc w/Enclosure:
Vice President Operations/Plant Manager
Wolf Creek Nuclear Operating Corporation
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 Corporation
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 66612-1590
Attorney General
120 S.W. 10th Avenue, 2nd Floor
Topeka, KS 66612-1597
Wolf Creek Nuclear Operating Corporation - 3 -
County Clerk
Coffey County Courthouse
110 South 6th Street
Burlington, KS 66839
Chief, Radiation and Asbestos
Control Section
Bureau of Air and Radiation
Kansas Department of Health and
Environment
1000 SW Jackson, Suite 310
Topeka, KS 66612-1366
Chief, Technological Hazards
Branch
FEMA, Region VII
9221 Ward Parkway
Suite 300
Kansas City, MO 64114-3372
Wolf Creek Nuclear Operating Corporation - 4 -
Electronic Distribution by RIV:
Regional Administrator (Elmo.Collins@nrc.gov)
Deputy Regional Administrator (Chuck.Casto@nrc.gov)
DRP Director (Dwight.Chamberlain@nrc.gov)
DRP Deputy Director (Anton.Vegel@nrc.gov)
DRS Director (Roy.Caniano@nrc.gov)
DRS Deputy Director (Troy.Pruett@nrc.gov)
Senior Resident Inspector (Chris.Long@nrc.gov)
Resident Inspector (Charles.Peabody@nrc.gov)
Site Secretary (Shirley.Allen@nrc.gov)
Branch Chief, DRP/B (Geoffrey.Miller@nrc.gov)
Senior Project Engineer, DRP/B (Rick.Deese@nrc.gov)
Public Affairs Officer (Victor.Dricks@nrc.gov)
Technical Support Branch (Chuck.Paulk@nrc.gov)
RITS Coordinator (Marisa.Herrera@nrc.gov)
Regional Counsel (Karla.Fuller@nrc.gov)
Congressional Affairs Officer (Jenny.Weil@nrc.gov)
OEMail Resource
Regional State Liaison Officer (Bill.Maier@nrc.gov)
NSIR/DRP/EP (Robert.Kahler@nrc.gov)
OEDO RIV Coordinator (Leigh.Trocine@nrc.gov)
ROPreports
File located: R:\_REACTORS\_WC\2009\WC 2009-004RP-CML.doc ADAMS ML093140803
SUNSI Rev Compl. 7Yes No ADAMS 7Yes No Reviewer Initials GM
Publicly Avail 7Yes No Sensitive Yes 7 No Sens. Type Initials GM
RI:DRP/ SRI:DRP/ SPE:DRP/ C:DRS/EB1 C:DRS/EB2
CAPeabody CMLong PJayroe RLKellar NFOKeefe
/RA - E/ /RA - E/ /RA/ /RA WSifre for/ /RA/
10/20/2009 10/20/2009 11/10/2009 11/05/2009 11/05/2009
C:DRS/OB C:DRS/PSB1 C:DRS/PSB2 C:DRP/
SGarchow MPShannon GEWerner GBMiller
/RA/ /RA Johnson for/ /RA/ /RA/
11/05/2009 11/05/2009 11/06/2009 11/10/2009
OFFICIAL RECORD COPY T=Telephone E=E-mail F=Fax
U.S. NUCLEAR REGULATORY COMMISSION
REGION IV
Docket: 50-482
License: NPF-42
Report: 05000482/2009004
Licensee: Wolf Creek Operating Corporation
Facility: Wolf Creek Generating Station
Location: 1550 Oxen Lane SE
Burlington, Kansas
Dates: July 1 through September 30, 2009
Inspectors: C. M. Long, Senior Resident Inspector
C. A. Peabody, Resident Inspector
P. A. Jayroe, Project Engineer, Project Branch B
G. W. Apger, Operations Engineer
S. M. Alferink, Reactor Inspector, Engineering Branch 2
J. P. Adams, Reactor Inspector, Engineering Branch 1, DRS
C. M. Ryan, Reactor Inspector, Engineering Branch 1, DRS
G. P. Tutak, Reactor Inspector, Engineering Branch 2, DRS
Approved By: G. B. Miller, Chief, Project Branch B
Division of Reactor Projects
-1- Enclosure
SUMMARY OF FINDINGS
IR 05000482/2009004, 7/1/2009 - 9/30/2009; Wolf Creek Generating Station, Integrated
Resident and Regional Report; Operability Evaluations; Post Maintenance Testing; Plant
Modifications; Maintenance Risk Assessments and Emergent Work Control; Fire Protection;
Event Followup.
The report covered a 3-month period of inspection by resident inspectors and an announced
baseline inspections by regional based inspectors. Seven Green and one Severity Level IV
noncited violations of significance were identified. The significance of most findings is indicated
by their color (Green, White, Yellow, or Red) using Inspection Manual Chapter 0609,
Significance Determination Process. Findings for which the significance determination
process does not apply may be Green or be assigned a severity level after NRC management
review. The NRC's program for overseeing the safe operation of commercial nuclear power
reactors is described in NUREG 1649, Reactor Oversight Process, Revision 4, dated
December 2006.
A. NRC-Identified Findings and Self-Revealing Findings
Cornerstone: Mitigating Systems
- Green. On June 30, 2009, the inspectors identified a noncited violation of Technical
Specification 3.8.1 for failure to perform an adequate common cause evaluation within
24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to demonstrate no common cause failure mechanism existed between the
emergency diesel generators after a through-wall leak was discovered on the essential
service water piping. Wolf Creek did not start the opposite train emergency diesel
generator and declared that the through-wall flaw was not a common cause failure
without any evaluation or supporting statements. Nondestructive testing had not been
started at this time. Subsequent evaluation of the flaw per American Society of
Mechanical Engineers (ASME) Code Case N513.2 restored operability to the essential
service water piping. The licensee entered this issue in their corrective action program
as Condition Report 18347.
The inspectors determined that the failure to demonstrate, per Technical
Specifications 3.8.1 Required Actions B.3.1 or B.3.2, that no common cause failure
existed for the emergency diesel generators 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
affected the cornerstone objective to ensure the availability, reliability, and capability of
systems that respond to initiating events to prevent undesirable consequences. The
inspectors evaluated the significance of this finding using Phase 1 of Inspection Manual
Chapter 0609, Appendix A, "Significance Determination of Reactor Inspection Findings
for At Power Situations," and determined that the finding was of very low safety
significance (Green) because the issue was not a design or qualification deficiency
confirmed to result in loss of operability or functionality, did not represent a loss of
system safety function, an actual loss of safety function of a single train for greater than
its technical specification allowed outage time, an actual loss of safety function of a
nontechnical specification risk-significant equipment train, and did not screen as
potentially risk significant due to a seismic, flooding, or severe weather initiating event.
-2- Enclosure
The cause of the finding has a problem identification and resolution crosscutting aspect
in the area associated with the corrective action program because Wolf Creek failed to
thoroughly evaluate the failure mechanism such that the resolutions address the causes
and extent of conditions, as necessary. Specifically Wolf Creek did not properly
consider the possibility of common-cause pitting failures which could have impacted the
essential service water piping Train A structural integrity thereby affecting its cooling
loads, including the Emergency Diesel Generator A P.1(c) (Section 1R15).
- Green. The inspectors identified a noncited violation of Technical Specification 3.8.1,
Required Action B.4.2.2 on March 24, 2009 when the licensee performed elective
maintenance on safety bus relays and removed equipment from service that was
required by the technical specification and the NRC Safety Evaluation Report (SER)
while in an extended diesel generator outage. The maintenance had the potential to
open the normal offsite feeder breaker. This issue has been entered into the corrective
action program as Condition Report 15727.
The inspectors determined that the failure to implement requirements of Technical
Specification 3.8.1 and the associated NRC safety evaluation 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 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 did not result in the Train B offsite power being inoperable for greater than
24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and did not involve external events such as flooding. Additionally, the cause of
the finding has a problem identification and resolution crosscutting aspect in the area
associated with the corrective action program. Specifically, Wolf Creek did an extent of
condition review in response to a previous violation which included
Procedure STS IC-208B, but still failed to prohibit performance of STS IC-208B during
the 7-day diesel outages P.1(c) (Section 1R19).
- Green. On August 22, 2009, the inspectors identified a noncited violation of Technical
Specification 3.0.3 in which both trains of Technical Specification 3.3.2 engineered safety
features actuation system interlock function 8.a were bypassed with jumper wires in
accordance with a plant procedure. Function 8.a is the interlock for reactor trip signal
coincident with lo Tave signal. Wolf Creek blocked the signal from the feedwater valves
with jumper wires during control rod drive motor-generator testing in Mode 3. The
inspectors and the NRR technical specification branch found this to be contrary to the
Updated Safety Analysis Report, the technical specifications, the technical specification
bases, and the NRC safety evaluations supporting the technical specifications. The
licensee entered this issue in their corrective action program as Condition Report 19318.
The inspectors found that the failure to implement Technical Specification 3.3.2 interlock,
function 8.a was a performance deficiency. The inspectors determined that this finding
was more than minor because it is associated with the design control attribute of the
Mitigating Systems Cornerstone and it affected the cornerstone objective to ensure the
availability, reliability, and capability of mitigating systems that respond to initiating
events to prevent undesirable consequences (i.e., core damage). The inspectors
evaluated the significance of this finding using Inspection Manual Chapter 0609.04,
-3- Enclosure
Phase 1 - Initial Screening and Characterization of Findings, and screened the finding
to Phase 2 because the finding represents a loss of a systems function. The inspectors
used Inspection Manual Chapter 0609, Appendix A and screened the finding to the NRC
senior reactor analyst for review because there was not an acceptable equipment
deficiency in the pre-solved worksheet. The senior reactor analyst determined that the
finding is Green because he solved Table 3.10 of the Risk-Informed Inspection
Notebook for Wolf Creek Generating Station, Revision 2.1a and found that the loss of
feedwater isolation signal for less than 3 days resulted in a 1E-7 (Green) outcome. The
inspectors also determined that the cause of the finding has a crosscutting aspect in the
human performance area associated with decision making because Wolf Creek failed to
make a risk significant decision using a systematic process. This issue was evaluated
more than once and those evaluations sought to justify bypassing the interlock rather
than seek the full regulatory basis for the interlock H.1.a] (1R15).
- Green. The inspectors identified a noncited violation of 10 CFR 50 Appendix B,
Criterion III, Design Control, for failing to translate the boric acid design basis into
procedures that ensure time sensitive operator actions are completed to achieve the
core shutdown margin specified in the core operating limits report. Performance
Improvement Request 2005-3461 identified that if the room coolers were started while
lake temperature was low, the boric acid solution temperature may decrease below the
solubility limit. Corrective actions for heat tracing and room temperature logging took
approximately 3 years to implement and stopped short of addressing boric acid system
operation when nonsafety power is lost to the heat tracing and the plant must be taken
to cold shutdown in accordance with technical specifications. The licensee entered this
issue in their corrective action program as Condition Report 20717.
The failure to translate the design bases into procedures that ensure the function of the
safety-related boric acid system upon loss of nonsafety-related heat tracing is a
performance deficiency. The inspectors determined that this finding was more than
minor because this issue aligned with Inspection Manual Chapter 0612, Appendix E,
example 2.f, because the pipe temperature was required to stay above the boric acid
solubility limit and the loss of the heat tracing and or room temperature decrease will
block the boric acid system. This issue was associated with the equipment performance
attribute of the mitigating systems cornerstone and affected the cornerstone objective to
ensure the availability, reliability, and capability of systems that respond to initiating
events. The inspectors evaluated the significance of this finding using Phase 1 of
Inspection Manual Chapter 0609, Appendix A, "Significance Determination of Reactor
Inspection Findings for At Power Situations," and determined that the finding screened to
phase 2 because the issue was a design or qualification deficiency confirmed to result in
loss of operability or functionality The inspectors evaluated the significance of this
finding using Phase 2 of Inspection Manual Chapter 0609, Risk Informed Inspection
Notebook for Wolf Creek Generating Station, and determined that the finding was of very
low safety significance because loss of the boric acid system in Table 3.9 for one year
resulted in a 1E-7 CDF when giving recovery credit for the refueling water storage tank.
The inspectors determined that this finding has a crosscutting aspect in the area of
problem identification and resolution associated with the corrective action program
component because Wolf Creek did not take appropriate corrective actions to resolve
known deficiencies in the design and operation of the boric acid system for
-4- Enclosure
approximately 4 years. The issue was re-evaluated in 2009, and the licensee failed to
correct the deficiencies identified in 2005. P.1.d] (Section 1R18).
- Green. The inspector identified a noncited violation of 10 CFR 50.65(a)(4) for failure to
adequately assess and manage the increase in risk during fuse inspection of component
cooling water valves supplying cooling loads inside containment. On March 18, 2009,
component cooling water Valves EG HV-16 and EG HV-54 were out of service for fuse
inspections to verify wiring for fire protection analyses. The inspectors observed that the
evolution was not included in the weekly risk assessment and that operations and
maintenance personnel did not have guidance or briefings for restoration of the valves.
Review of the risk assessment revealed that the impact of de-energizing the valves in
the closed position was neglected and that restoration actions credited by the risk
analyst were unknown to the control room and craft workers. The issue was entered into
the corrective action program as Condition Report 15318.
The failure to adequately assess and manage risk in accordance with AP 22C-003 and
the preplanned risk assessment for the use of local actions to ensure component cooling
water cooling to loads inside containment was a performance deficiency. The finding is
more than minor because the licensee failed to effectively manage prescribed significant
compensatory measures for maintenance activities that could increase the likelihood of
initiating events. The finding was of very low safety significance because the magnitude
of the calculated risk deficit was less than IE-6 even though risk management actions
were not in place. The inspectors also determined that the finding has a human
performance crosscutting aspect in the area associated with work control because the
risk assessment procedure and clearance order procedure assumed local actions could
be accomplished but there was no communication regarding this during the work
planning stages H.3(b) (Section 1R13).
- Severity Level IV. The inspectors identified a Severity Level IV noncited violation of
License Condition 2.C.(5), Fire Protection, for making changes to the approved fire
protection program without the required prior Commission approval. Specifically, the
licensee made a change to the Updated Safety Analysis Report that allowed the
licensee to violate the requirements of 10 CFR Part 50, Appendix R, Section III.L.
Specifically, when the licensee recognized that fire damage could cause a pressurizer
power operated relief valve to open long enough to create a void in the reactor vessel,
this was documented as acceptable when it was not in compliance with this regulatory
requirement. The licensee entered this issue into their corrective action program as
Performance Improvement Request 2008-004869.
This finding was assessed using traditional enforcement since it had the potential for
impacting the NRCs ability to perform its regulatory function. This finding is more than
minor since the change required prior staff review and approval prior to implementation
and it did not receive the required approval. A senior reactor analyst performed a
Phase 3 evaluation and determined this performance deficiency was of very low risk
significance. In accordance with the guidance in Supplement I of the Enforcement
Policy, this issue is considered a Severity Level IV noncited violation because it is of
very low risk significance. This finding had a crosscutting aspect in the area of human
performance associated with resources because the licensee failed to maintain
long-term plant safety by maintaining design margins. Specifically, the licensees choice
-5- Enclosure
to allow reactor vessel head voiding during an alternative shutdown in lieu of restoring
the plant to compliance with the requirements of 10 CFR Part 50, Appendix R,
Section III.L constituted a reduction in safety margin H.2(a) (Section 40A5.3).
Cornerstone: Barrier Integrity
- Green. The inspectors identified a noncited violation of Technical Specification 5.4.1.a,
Procedures, for failure to follow Procedure AP 12-003, Foreign Material Exclusion.
On August 12, 2009, the inspectors conducted a walkdown of the spent fuel pool area
and found duct tape attached to various fueling and control rod tools such that duct tape
was below the water. This duct tape was not in the foreign material exclusion logs.
Spent fuel pool foreign material control is required under Procedure AP 12-003. The
licensee entered this issue in their corrective action program as Condition Report 20338.
The inspectors determined that the failure to log material in accordance with
Procedure AP 12-003 was a performance deficiency. This finding is more than minor
because it impacted the Barrier Integrity Cornerstone attribute of configuration control
and affected the cornerstone objective to maintain functionality of the spent fuel pool
system. Using Inspection Manual Chapter 0609.04, Phase 1 - Initial Screening and
Characterization of Findings, this finding was determined to be of very low safety
significance because the finding only affected the barrier function of the spent fuel pool.
The inspectors determined that this finding has a crosscutting aspect in the area of
problem identification and resolution associated with the corrective action program
component because although Wolf Creek performed a root cause and extent of
condition evaluation for untracked foreign material, the evaluation still failed to find the
duct tape in the pool itself. This allowed the tape to continue to be untracked P.1.c]
(Section 1R05).
Cornerstone: Miscellaneous
- Severity Level IV. The inspectors identified a Severity Level IV noncited violation of
10 CFR 50.73, Licensee Event Report System, with three examples in which the
licensee failed to submit licensee event reports within 60 days following discovery of an
event meeting the reportability criteria. First, on April 10, 2008, Wolf Creek submitted
Licensee Event Report 2008-002-00 under 10 CFR 50.73(a)(2)(i)(B) which is operation
prohibited by technical specifications but failed to make a report for a loss of safety
function per 10 CFR 50.73(a)(2)(v) for the same event in which both trains of the
emergency core cooling system were inoperable on February 13-14, 2008. Second,
Wolf Creek filed Licensee Event Report 2008-004-00 on June 6, 2008 under
50.73(a)(2)(iv)(A) for an event that caused automatic start of an emergency diesel during
a loss of offsite power on April 16, 2008. No report was made under 50.73(a)(2)(v) for
an event or condition that could have prevented a safety function due to the loss of
offsite power. Third, on April 10, 2008, Wolf Creek filed Event Notification Report 44131
under 10 CFR 50.72(b)(3)(ii)(B) based on a possible trip of all four containment coolers.
The notification was later retracted. The inspectors found insufficient evidence to show
that the containment coolers would not trip and concluded the event should have been
reported under 10 CFR 50.73(a)(2)(v). All three issues are collectively captured in
Condition Report 15318.
-6- Enclosure
The inspectors reviewed this issue in accordance with Inspection Manual Chapter 0612
and the NRC Enforcement Manual. Through this review, the inspectors determined that
traditional enforcement was applicable to this issue because the NRC's regulatory ability
was affected. Specifically, the NRC relies on the licensee to identify and report
conditions or events meeting the criteria specified in regulations in order to perform its
regulatory function, and when this is not done, the regulatory function is impacted. The
inspectors determined that this finding was not suitable for evaluation using the
significance determination process, and as such, was evaluated in accordance with the
NRC Enforcement Policy. The finding was reviewed by NRC management, and because
the violation was determined to be of very low safety significance, was not repetitive or
willful, and was entered into the corrective action program, this violation is being treated
as a Severity Level IV noncited violation consistent with the NRC Enforcement Policy.
This finding was determined to have a crosscutting aspect in the area of problem
identification and resolution associated with the corrective action program in that the
licensee failed to appropriately and thoroughly evaluate for reportability aspects all
factors and time frames associated with the inoperability of the emergency core cooling
system, the offsite power system, and the containment heat removal system P.1(c)
(Section 4OA3).
-7- Enclosure
REPORT DETAILS
Summary of Plant Status
The plant started the inspection period at 100 percent rated thermal power. On August 19,
2009, Wolf Creek experienced and automatic reactor trip from 100 percent power when a
lightning strike caused a loss of offsite power. Wolf Creek restarted on August 24, 2009. On
August 28, 2009, Wolf Creek reduced power to 99 percent for the end of core life moderator
temperature coefficient surveillance test. On September 30, 2009, Wolf Creek decreased to
97 percent power for heater drain pump repair.
1. REACTOR SAFETY
Cornerstones: Initiating Events, Mitigating Systems, Barrier Integrity, and
1R01 Adverse Weather Protection (71111.01)
.1 Summer Readiness for Offsite and Alternate ac Power
a. Inspection Scope
On August 10, 2009, the inspectors performed a review of the licensees preparations for
summer weather for selected systems, including conditions that could lead to loss-of-
offsite power and conditions that could result from high temperatures. The inspectors
reviewed the licensees procedures affecting these areas and the communications
protocols between the transmission system operator and the plant to verify that the
appropriate information was being exchanged when issues arose that could affect the
offsite power system. Examples of aspects considered in the inspectors review
included:
- The coordination between the transmission system operator and the plant during
offnormal or emergency events
- The explanations for the events
- The estimates of when the offsite power system would be returned to a normal
state
- The notifications from the transmission system operator to the plant when the
offsite power system was returned to normal
These activities constitute completion of one readiness for summer weather affect on
offsite and alternate ac power sample as defined in Inspection
Procedure (IP) 71111.01-05.
b. Findings
No findings of significance were identified.
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1R04 Equipment Alignments (71111.04)
.1 Partial Walkdown
a. Inspection Scope
The inspectors performed partial walkdown of the following risk-significant systems:
- Motor-Driven auxiliary feedwater Train B, July 7, 2009
- Turbine-Driven auxiliary feedwater, July 7, 2009
- Essential service water Train A, August 9, 2009
- Centrifugal charging pump Train A, September 22, 2009
The inspectors selected these systems based on their risk-significance relative to the
Reactor Safety Cornerstones at the time they were inspected. The inspectors attempted
to identify any discrepancies that could affect the function of the system, and, therefore,
potentially increase risk. The inspectors reviewed applicable operating procedures,
system diagrams, Updated Safety Analysis Report (USAR), technical specification
requirements, administrative technical specifications, outstanding work orders, condition
reports, and the impact of ongoing work activities on redundant trains of equipment in
order to identify conditions that could have rendered the systems incapable of
performing their intended functions. The inspectors also 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. Specific
documents reviewed during this inspection are listed in the attachment.
These activities constitute completion of four partial system walkdown sample as defined
in IP 71111.04-05.
b. Findings
No findings of significance were identified.
.2 Complete Walkdown
a. Inspection Scope
On September 18, 2009, the inspectors performed a complete system alignment
inspection of the main steam system to verify the functional capability of the system.
The inspectors selected this system because it was considered both safety significant
and risk significant in the licensees probabilistic risk assessment. The inspectors
walked down the system to review mechanical and electrical equipment line ups,
-9- Enclosure
electrical power availability, system pressure and temperature indications, as
appropriate, component labeling, component lubrication, component and equipment
cooling, hangers and supports, operability of support systems, and to ensure that
ancillary equipment or debris did not interfere with equipment operation. The inspectors
reviewed a sample of past and outstanding work orders to determine whether any
deficiencies significantly affected the system function. In addition, the inspectors
reviewed the corrective action program database to ensure that system equipment-
alignment problems were being identified and appropriately resolved. Specific
documents reviewed during this inspection are listed in the attachment.
These activities constitute completion of one complete system walkdown sample as
defined in IP 71111.04-05.
b. Findings
No findings of significance were identified.
1R05 Fire Protection (71111.05)
.1 Quarterly Fire Inspection Tours
a. Inspection Scope
The inspectors conducted fire protection walkdowns that were focused on availability,
accessibility, and the condition of firefighting equipment in the following risk-significant
plant areas:
- 4kV Switchgear rooms, control building 2000 elevation, July 9, 2009
- Diesel generator rooms, diesel building 2000 elevation, July 9, 2009
- Turbine-Driven auxiliary feedwater room, August 11, 2009
- Spent fuel pool 2047 elevation, August 12, 2009
The inspectors reviewed areas to assess if licensee personnel had implemented a fire
protection program that adequately controlled combustibles and ignition sources within
the plant; effectively maintained fire detection and suppression capability; maintained
passive fire protection features in good material condition; and had implemented
adequate compensatory measures for out of service, degraded or inoperable fire
protection equipment, systems, or features, in accordance with the licensees fire plan.
The inspectors selected fire areas based on their overall contribution to internal fire risk
as documented in the plants individual plant examination of external events with later
additional insights, their potential to affect equipment that could initiate or mitigate a plant
transient, or their impact on the plants ability to respond to a security event. Using the
documents listed in the attachment, the inspectors verified that fire hoses and
extinguishers were in their designated locations and available for immediate use; that
fire detectors and sprinklers were unobstructed, that transient material loading was
within the analyzed limits; and fire doors, dampers, and penetration seals appeared to
be in satisfactory condition. The inspectors also verified that minor issues identified
during the inspection were entered into the licensees corrective action program.
Specific documents reviewed during this inspection are listed in the attachment.
- 10 - Enclosure
These activities constitute completion of four quarterly fire protection inspection samples
as defined by IP 71111.05-05.
b. Findings
.2 Introduction. On August 12, 2009, the inspectors identified a Green noncited violation of
Technical Specification 5.4.1.a, Procedures, for failure to follow AP 12-003, Foreign
Material Exclusion, after a root cause assessment on foreign material exclusion.
Description. On August 12, 2009, the inspectors conducted a walkdown of the spent fuel
pool area and found duct tape below the water. Numerous pieces of duct tape, including
that on fueling and control rod manipulation tools, were not in the logs. The inspectors
reviewed Procedure AP 12-003, Foreign Material Exclusion, Revision 7. The area
surrounding the spent fuel pool is posted as a foreign material exclusion area, a
contaminated area, and a hot particle area. Procedure AP 12-003 requires the highest
level of foreign material accountability, or Level 1, for the spent fuel pool. Level 1
requires several actions: All materials in the area are to be described; all materials are
logged in and out; logs specify how material was removed; logs identify the person
writing on the log itself; and the pages of the log itself are tracked. The inspectors
reviewed the spent fuel pool area logs and concluded the logs were inadequate.
Although Wolf Creek logged some duct tape, numerous pieces of duct tape on fuel and
control rod tools were not logged.
The inspectors reviewed the spent fuel pool area material tracking practices since the
completion of a root cause evaluation and extent of condition review in response to
previous NRC finding 05000482/2009002-03. The inspectors found that Wolf Creek
performed an extent of condition review to examine the bottom of the cask pit and the
spent fuel racks, but failed to identify the duct tape on the tools. The inspectors did not
find any documentation stating that the tape was acceptable for use underwater in an
acidic environment. Although the tape markings are used for refueling operations, the
inspectors found no documentation that would lead Wolf Creek to identify the missing
tape. On August 12, 2009, Wolf Creek initiated Condition Report 19110, but this report
only asked how to handle the submerged tape and did not identify the failure to log the
material. The issue was appropriately captured in the corrective action program with
Condition Report 20338.
Analysis. The inspectors determined that the failure to track foreign material in
accordance with Procedure AP 12-003 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. This finding is more
than minor because it impacted the Barrier Integrity Cornerstone attribute of
configuration control and affected the cornerstone objective to maintain functionality of
the spent fuel pool system. Using Manual Chapter 0609.04, Phase 1 - Initial Screening
and Characterization of Findings, this finding was determined to be of very low safety
significance because the finding only affected the barrier function of the spent fuel pool
and did not result in actual clogging of the system. The inspectors determined that this
finding has a crosscutting aspect in the area of problem identification and resolution
associated with the corrective action program component because although Wolf Creek
- 11 - Enclosure
performed a root cause and extent of condition evaluation for untracked foreign material,
the evaluation still failed to find the duct tape in the pool itself. This allowed the tape to
continue to be untracked P.1.c].
Enforcement. Technical Specification 5.4.1.a requires the implementation of written
procedures described in Regulatory Guide 1.33, Revision 2, Appendix A, including
procedures for performing maintenance that can affect the performance of safety-related
equipment. Procedure AP 12-003, Foreign Material Exclusion, Revision 6, requires
foreign material accountability for the spent fuel pool. Contrary to the above, prior to
August 12, 2009, the licensee failed account for foreign material in the spent fuel pool.
Because this violation was determined to be of very low safety significance and was
placed in the corrective action program as Condition Report 20338, this violation is being
treated as a noncited violation in accordance with Section VI.A.1 of the Enforcement
Policy: NCV 05000482/2009004-01, Failure to Log Foreign Material in Spent Fuel Pool
after Extent of Condition Evaluation.
1R06 Flood Protection Measures (71111.06)
a. Inspection Scope
The inspectors reviewed the USAR, the flooding analysis, and plant procedures to
assess seasonal susceptibilities involving internal flooding; reviewed the USAR and
corrective action program to determine if licensee personnel identified and corrected
flooding problems; inspected underground bunkers/manholes to verify the adequacy of
sump pumps, level alarm circuits, cable splices subject to submergence, and drainage
for bunkers/manholes; verified that operator actions for coping with flooding can
reasonably achieve the desired outcomes; and walked down the one area listed below to
verify the adequacy of equipment seals located below the flood line, floor and wall
penetration seals, watertight door seals, common drain lines and sumps, sump pumps,
level alarms, and control circuits, and temporary or removable flood barriers. Specific
documents reviewed during this inspection are listed in the attachment.
- September 17, 2009, Essential service water Manhole MHE-2B for cable splice
inspections.
These activities constitute completion of one flood protection measures inspection
sample as defined by IP 71111.06-05.
b. Findings
No findings of significance were identified.
1R11 Licensed Operator Requalification Program (71111.11)
.1 Quarterly Inspection
a. Inspection Scope
On September 14, 2009, the inspectors observed a crew of licensed operators in the
plants simulator to verify that operator performance was adequate, evaluators were
- 12 - Enclosure
identifying and documenting crew performance problems, and training was being
conducted in accordance with licensee procedures. The inspectors evaluated the
following areas:
- Licensed operator performance
- Crews clarity and formality of communications
- Crews ability to take timely actions in the conservative direction
- Crews prioritization, interpretation, and verification of annunciator alarms
- Crews correct use and implementation of abnormal and emergency procedures
- Control board manipulations
- Oversight and direction from supervisors
- Crews ability to identify and implement appropriate technical specification
actions and emergency plan actions and notifications
The inspectors compared the crews performance in these areas to pre-established
operator action expectations and successful critical task completion requirements.
Specific documents reviewed during this inspection are listed in the attachment.
These activities constitute completion of one quarterly licensed-operator requalification
program sample as defined in IP 71111.11.
b. Findings
No findings of significance were identified.
.2 Annual Inspection (71111.11B)
The licensed operator prequalification program involves two training cycles that are
conducted over a 2-year period. In the first cycle, the annual cycle, the operators are
administered an operating test consisting of job performance measures and simulator
scenarios. In the second part of the training cycle, the biennial cycle, operators are
administered an operating test and a comprehensive written examination.
a. Inspection Scope
The inspector conducted an in office review of the annual prequalification training
program operating test results for 2009. The licensee examined fifty operators (twenty-
one reactor operators and twenty-nine senior reactor operators) during this
prequalification cycle. In addition, nine operating crews were examined on the facility's
simulator. All of the operating crews passed the simulator scenarios and all operators
passed the operating tests.
- 13 - Enclosure
b. Findings
No findings of significance were identified.
1R13 Maintenance Risk Assessments and Emergent Work Control (71111.13)
a. Inspection Scope
The inspectors reviewed licensee personnel's evaluation and management of plant risk
for the maintenance and emergent work activities affecting risk-significant and safety-
related equipment listed below to verify that the appropriate risk assessments were
performed prior to removing equipment for work:
- Performance of Procedure TMP 09-014, July 15, 2009
- Failure of flow indicator BBFI-425, July 16, 2009
- Component cooling water valves fuse inspections, March 16, 2009
- Week of August 24, 2009, planned work risk assessment
The inspectors selected these activities based on potential risk-significance relative to
the Reactor Safety Cornerstones. As applicable for each activity, the inspectors verified
that licensee personnel performed risk assessments as required by 10 CFR 50.65(a)(4)
and that the assessments were accurate and complete. When licensee personnel
performed emergent work, the inspectors verified that the licensee personnel promptly
assessed and managed plant risk. The inspectors reviewed the scope of maintenance
work, discussed the results of the assessment with the licensee's probabilistic risk
analyst or shift technical advisor, and verified plant conditions were consistent with the
risk assessment. The inspectors also reviewed the technical specification requirements
and inspected portions of redundant safety systems, when applicable, to verify risk
analysis assumptions were valid and applicable requirements were met. Specific
documents reviewed during this inspection are listed in the attachment.
These activities constitute completion of four maintenance risk assessments and
emergent work control inspection samples as defined by IP 71111.13-05.
b. Findings
Introduction. The inspector identified a noncited violation of 10 CFR 50.65(a)(4) for
failure to adequately assess and manage the increase in risk during fuse inspection of
component cooling water valves supplying cooling loads inside containment.
Description. On March 18, 2008, component cooling water Valves EG HV-16 and EG
HV-54 were out of service for fuse inspections to verify wiring for fire protection
analyses. The inspectors observed that the evolution was not included in the weekly risk
assessment. The inspectors noted that operations and maintenance personnel did not
have guidance or briefings for restoration of the valves. Review of the risk assessment
by Wolf Creek after inspector questioning revealed that the impact of de-energizing the
valves in the close position was neglected and that restoration actions credited by the
risk analyst were unknown to the control room and craft workers. Specifically, Condition
Report 0015318 states that loss of reactor coolant pump thermal barriers was possible,
- 14 - Enclosure
but it did not state that it was possible to also lose the reactor coolant pump seal water
cooling heat exchanger, reactor coolant pump radial bearing cooling, reactor coolant
pump motor cooling, the letdown heat exchanger, and the excess letdown heat
exchanger. Due to these additional loads, this would also be a reactor trip initiator
initiating event.
A local clearance order was assumed to provide local control and restoration instructions
but Wolf Creek later found that these actions were not possible because the clearance
order contained no restoration instructions. Inspectors reviewed the work package for
the fuse inspections and found that the pre-job briefing did not contain any instructions to
craftsmen for rapid restoration. Procedure AP 22C-003, Revision 13 Attachment A
required risk management actions due to the potential for a trip initiator and the potential
to interrupt the thermal barrier heat exchangers for the reactor coolant pump seals. The
control room and maintenance personnel were not aware of the restoration actions
assumed in the risk assessment. Only the risk engineers were aware of the restoration
actions. Wolf Creeks evaluation of the issue found that the maintenance planning
group, the risk assessment engineers, and operations were not procedurally required to
discuss the restoration actions when changes were made during maintenance planning
in the prior weeks.
Analysis. The failure to adequately assess and manage risk in accordance with
AP 22C-003 and the preplanned risk assessment for the use of local actions to ensure
component cooling water cooling to loads inside containment 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 finding is more than minor because the licensee failed to effectively manage
prescribed significant compensatory measures for maintenance activities that could
increase the likelihood of initiating events. The finding was of very low safety significance
because the magnitude of the calculated risk deficit was less than 1 x 10-6 even though
risk management actions were not in place. The inspectors also determined that the
finding has a human performance crosscutting aspect in the area associated with work
control because the risk assessment procedure and clearance order procedure assumed
local actions could be accomplished but there was no communication regarding this
during the work planning stages H.3(b)
Enforcement. 10 CFR 50(a)(4), requires, in part, that before performing maintenance
activities (including but not limited to surveillance, post maintenance testing, and
corrective and preventive maintenance), the licensee shall assess and manage the
increase in risk that may result from the proposed maintenance activities.
Procedure AP 22C-003, Revision 13, and the resulting weekly risk assessment
implement this regulation. Contrary to the above, on March 18, 2009, the licensee did
not effectively manage the increase in risk resulting from a maintenance activity.
Specifically, on March 18, 2009, during fuse inspections of component cooling water
Valves EG HV-16 and EG HV-54, the licensee failed to adequately assess and manage
the increase in risk that resulted from the maintenance activity. Restoration actions
credited in Wolf Creeks weekly risk assessment were determined to be not possible to
implement. The licensee entered this issue into their corrective action program as
Condition Report 15318. Because the licensee has entered the issue into their corrective
- 15 - Enclosure
action program and the finding is of very low safety significance, this violation is being
treated as a NCV, consistent with Section VI.A of the NRC Enforcement Policy:
NCV 5000482/2009004-02, Inability to perform manual actions for risk assessment.
1R15 Operability Evaluations (71111.15)
a. Inspection Scope
The inspectors reviewed the following issues:
- Essential service water piping through wall leakage, separate occurrences on
June 30, July 28, and August 19, 2009
- Diesel generator common cause failure evaluation on June 30, 2009
- Performance of procedure SYS SB-122 on August 22, 2009
- Nonconservative core flux technical specification on August 5, 2009
The inspectors selected these potential operability issues based on the risk-significance
of the associated components and systems. The inspectors evaluated the technical
adequacy of the evaluations to ensure that technical specification operability was
properly justified and the subject component or system remained available such that no
unrecognized increase in risk occurred. The inspectors compared the operability and
design criteria in the appropriate sections of the technical specifications and USAR to
the licensees evaluations, to determine whether the components or systems were
operable. Where compensatory measures were required to maintain operability, the
inspectors determined whether the measures in place would function as intended and
were properly controlled. The inspectors determined, where appropriate, compliance
with bounding limitations associated with the evaluations. Additionally, the inspectors
also reviewed a sampling of corrective action documents to verify that the licensee was
identifying and correcting any deficiencies associated with operability evaluations.
Specific documents reviewed during this inspection are listed in the attachment.
These activities constitute completion of four operability evaluations inspection samples
as defined in IP-1111.15-05
b. Findings
.1 Introduction. The inspectors identified a Green noncited violation of Technical
Specification 3.8.1 for failure to perform an adequate common cause evaluation within
24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to demonstrate no common cause failure mechanism existed between the
operable and inoperable emergency diesel generators.
Description. At 11:15 a.m., on June 30, 2009, Wolf Creek auxiliary building watch
discovered a through-wall leak in the essential service water Train B
Piping EF-138-HBC-30 just upstream of valve EF-HV-0038. The piping was leaking
through two adjacent pinholes at the bottom of the pipe spaced approximately 0.4 inch
apart. This condition was recognized as a limiting condition of operations per
Condition A of Technical Requirements Manual 3.4.17, Structural Integrity, which
- 16 - Enclosure
requires the structural integrity of all ASME Class I, II, and III piping to be maintained.
The required action directed operators to declare the essential service water Train B
inoperable. Thus Wolf Creek entered Condition A of Technical Specification 3.7.8
Essential Service Water, for one train of essential service water inoperable. This
condition has a required action of restoring the essential service water train to operable
status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, but it also requires simultaneous entry into Condition B of
Technical Specification 3.8.1, AC Sources Operating, for the emergency diesel
generator made inoperable by the essential service water system. There are four
required actions associated with Technical Specification 3.8.1, Condition B. First,
Required Action B.1, the control room operators are to verify correct breaker alignment
and indicated power availability for each offsite power circuit within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and every
8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> thereafter. Second, Required Action B.2 requires that features supported by the
inoperable diesel generator be declared inoperable when its required redundant feature
is inoperable within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. Third, Required Action B.3.1 requires Wolf Creek to
determine that the operable diesel generator is not inoperable due to a common cause
failure. Alternatively, Required Action B.3.2 directs Wolf Creek to verify the operable
diesel generator starts from standby conditions and achieves steady state voltage and
frequency, within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. Fourth, Required Action B.4.1 directs the restoration the
diesel generator to operable status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. Wolf Creek properly carried out
Required Actions B.1, B.2, and B.4.1 required by Technical Specification 3.8.1,
Condition B.
At 12:02 p.m., 47 minutes after the leak was discovered, the control room logs state that
Technical Specification 3.8.1, Action B.3.1, is being exited because Emergency Diesel
Generator B inoperable due to ESW being inoperable not a common cause failure. The
inspectors interviewed operations personnel on the adequacy of such a justification.
Operations provided the inspectors with a completed copy of Procedure SYS KJ-200,
Inoperable Emergency Diesel. Procedure Step 6.1.5 states: If the absence of any
potential common cause failure can be demonstrated . . . then document the evaluation
on the cover sheet. However, the cover sheet had only one sentence which matched
the log entry verbatim. At the time of this determination, ultrasonic testing to determine
flaw size and pipe wall thicknesses had yet to be performed. The results of that testing
were the basis for an ASME N513.2 code case which eventually restored operability.
During later interviews regarding the control room log entries, Wolf Creek stated that
nonlicensed operators did not find any other through wall leaks on essential service
water Train B, and therefore Train B was operable. The inspectors found that this type
of visual evaluation did not meet the reasonable assurance standard specified in RIS 2005-20. Visual examinations can not identify below minimum wall thickness piping or
piping flaws under insulation. The inspectors concluded the licensees evaluation lacked
a valid technical basis for determination that a common cause failure mechanism did not
exist on the opposite train emergency diesel generator.
The ASME N513.2 code case was issued and essential service water/emergency diesel
generator operability restored at 9:40 p.m. that night. The code case verified the
structural integrity of the piping despite the current through-wall flaw; however, it
specified that due to the potential common cause nature of pitting flaws, five additional
locations had to be ultrasonic tested to verify that minimum wall thickness was met.
Although none of the additional locations indicated any below minimum-wall flaws in the
essential service water piping, an expanded ultrasonic test of the leak area revealed two
- 17 - Enclosure
additional pits that were below the minimum wall thickness acceptance criteria.
Separate evaluations were performed for those flaws and all three were permanently
repaired per the ASME code on July 23, 2009.
Analysis: The inspectors determined that the failure to demonstrate operability of
Emergency Diesel Generator B per Technical Specification 3.8.1, Required Action B.3.1
or B.3.2 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 latent common mode failure mechanism was not correctly evaluated. The inspectors
evaluated the significance of this finding using Phase 1 of Inspection Manual
Chapter 0609, Appendix A, "Significance Determination of Reactor Inspection Findings
for At Power Situations," and determined that the finding was of very low safety
significance (Green) because the issue was not a design or qualification deficiency
confirmed to result in loss of operability or functionality, did not represent a loss of
system safety function, an actual loss of safety function of a single train for greater than
its technical specification allowed outage time, an actual loss of safety function of a
nontechnical specification risk-significant equipment train, and did not screen as
potentially risk significant due to a seismic, flooding, or severe weather initiating event.
The cause of the finding has a problem identification and resolution crosscutting aspect
in the area associated with the corrective action program because Wolf Creek failed to
thoroughly evaluate the failure mechanism such that the resolutions address the causes
and extent of conditions, as necessary. Specifically, Wolf Creek did not properly
consider the possibility of common-cause pitting failures which could have impacted the
essential service water Train A piping structural integrity thereby affecting its cooling
loads, including Emergency Diesel Generator A (P.1(c)).
Enforcement: Technical Specification 3.8.1 Required Actions B.3.1 and B.3.2 require,
with one diesel generator inoperable, to determine that the operable diesel generator is
not inoperable due to common cause failure or else perform SR 3.8.1.2 [run the diesel
generator]. Contrary to this requirement, on June 30, 2009, the licensee failed to
demonstrate that Emergency Diesel Generator A was operable by evaluation of common
cause failure or by performing SR 3.8.1.2 while emergency diesel generator B was
inoperable due to essential service water piping corrosion. Specifically, the control room
logs exited Required Action B.3.1 stating that EDG B inoperable due to ESW being
inoperable not a common cause failure. No further evaluation was provided. Because
the finding is of very low safety significance and has been entered into the corrective
action program as Condition Report 18347, this violation is being treated as an noncited
violation, consistent with Section VI.A of the NRC Enforcement Policy:
NCV 05000482/2009004-03: Inadequate Evaluation of Emergency Diesel Generator for
Common Cause Failure in the Supporting Essential Service Water System.
- 18 - Enclosure
.2 Introduction. On August 22, 2009, the inspectors identified a violation of Technical
Specification 3.0.3 in which both trains of a Technical Specification 3.3.2 interlock in the
engineered safety features actuation system were bypassed with jumper wires in
accordance with plant procedure.
Description. On August 22, 2009, the inspectors observed that both trains of Technical
Specification 3.3.2, function 8.a, P-4, were bypassed while in Mode 3. The inspectors
found that Wolf Creek installed jumper wires on both trains in accordance with
Procedure SYS SB-122, Enabling/Disabling P-4/Lo Tave FWIS [feed water isolation
signal]. The inspectors found that Wolf Creek has installed the jumper wires on both
trains in the past to support reactor trip breaker and control rod drop testing in Mode 3.
The jumpers defeated the function of both trains of reset switches on the main control
board such that a P4/FWIS cannot be sent to close feedwater valves and trip the main
feedwater pumps.
The inspectors reviewed the technical specification bases for the engineered safety
features actuation system interlocks and function 8.a. The bases and USAR state that
the functions of the interlock are to: 1) trip the main turbine, 2) isolate main feed water
coincident with lo Tavg, 3) allow manual block of the automatic re-actuation of safety
injection after a manual reset of safety injection, 4) allow arming of the steam dump
valves and transfer the steam dump from the load rejection Tavg controller to the plant
trip controller, 5) prevents opening of the main feed water isolation valves if they were
closed on safety injection or steam generator hi-hi water level. The inspectors found that
this was consistent with the standard improved technical specifications for Westinghouse
plants and the Wolf Creek USAR, Table 7.3-15, NSSS Interlocks for Engineered Safety
Feature Actuation System. Under License Amendment 123, Wolf Creek converted to
improved standard technical specifications in December 1999. The P-4 interlock
description has not changed since 1999. The licensee submittals acknowledged that the
functions of P-4 were not part of a design basis analysis, but were retained in the
technical specifications to limit reactor coolant system cooldown following a reactor trip.
Technical Specification 3.3.2 states that The ESFAS [engineered safety features
actuation signal] instrumentation for each Function in Table 3.3.2 shall be OPERABLE
According to Table 3.3.2-1. Function 8 of Table 3.2.-1 covers interlocks and specifically
interlock 8.a, P-4, is required to be Operable in Modes 1, 2, and 3. The inspectors found
that function 8.a is required in Modes 1, 2, and 3. The inspectors consulted with the
Office of Nuclear Reactor Regulations technical specification branch and found that
statements in the bases provide a summary of the technical specification and do not
override requirements. The sentence in the bases that states: This Function must be
OPERABLE in MODES 1, 2, and 3 when the reactor may be critical or approaching
criticality, clarifies why it is required in Modes 1, 2, and 3 and does not permit P-4 to be
inoperable if the reactor is not approaching criticality. Operators are trained to anticipate
criticality such as during control rod-drive motor-generator testing during August 22-23,
2009.
During interviews, Wolf Creek stated that it was necessary to bypass the P4/FWIS in
order to perform rod-drive motor-generator set testing that cycled the reactor trip
breakers. Wolf Creek contended that the P-4/FWIS was not necessary to assure
compliance with the plant safety analysis. Lastly, Wolf Creek stated that during Mode 3
after refueling outages, it was necessary to install jumpers and bypass the P-4/FWIS for
- 19 - Enclosure
rod-drop testing because operation of the main feedwater system in automatic level
control was more desirable than having an operator manually control steam generator
levels with auxiliary feedwater. The inspectors agreed that this interlock is not assumed
in Chapter 15 of the USAR, but the inspectors found that the Wolf Creek technical
specification bases state that ESFAS instrumentation satisfies Criterion 3 of 10 CFR 50.36(c)(2)(ii) which is identical to the generic standard specifications approved by the
NRC. The inspectors found that there are several technical specification systems such
as steam generator atmospheric relief valves, the condensate storage tank, and
pressurizer power operated relief valves that are not in Chapter 15 of the USAR but are
required to be operable under technical specifications per 10 CFR 50.36. Thus, the
inspectors found that the interlocks absence in Chapter 15 of the USAR does not mean
it is not required by the technical specification. Wolf Creek previously evaluated this
condition in Performance Improvement Request 2001-0041 which concluded this
P-4/FWIS was not required to be operable in any Mode because it is not credited in
Chapter 15 of the USAR. Wolf Creek also used other plants with NRC approved safety
evaluations to justify the use of Procedure SYS SB-122 rather than requesting a license
amendment. The inspectors found that these conclusions are incorrect.
The inspectors found that control room operators did not log the inoperability of P-4 until
after inspector questioning, and afterward, operators incorrectly applied Technical Specification 3.3.2, Condition F, which allowed 60 hours6.944444e-4 days <br />0.0167 hours <br />9.920635e-5 weeks <br />2.283e-5 months <br /> to return one train of the
interlock to service. With both trains of P4 bypassed, Technical Specification 3.0.3
applied and Wolf Creek had 13 hours1.50463e-4 days <br />0.00361 hours <br />2.149471e-5 weeks <br />4.9465e-6 months <br /> to be in Mode 4. The P-4 interlock was inoperable
for approximately 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br /> from August 22-23, 2009. Wolf Creek missed the transition
to Mode 4.
Analysis. The inspectors found that the failure to evaluate implement Technical Specification 3.3.2 interlock, function 8.a was a performance deficiency. The inspectors
determined that this finding was more than minor because it is associated with the
design control attribute of the Mitigating Systems Cornerstone and it affected the
cornerstone objective to ensure the availability, reliability, and capability of mitigating
systems that respond to initiating events to prevent undesirable consequences (i.e., core
damage). The inspectors evaluated the significance of this finding using Inspection
Manual Chapter 0609.04, Phase 1 - Initial Screening and Characterization of Findings,
and screened the finding to Phase 2 because the finding represents a loss of a systems
function. The inspectors used Inspection Manual Chapter 0609, Appendix A and
screened the finding to the NRC senior reactor analyst for review because there was not
an acceptable equipment deficiency in the pre-solved worksheet. The senior reactor
analyst determined that the finding is Green because he solved Table 3.10 of the
Risk-Informed Inspection Notebook for Wolf Creek Generating Station, Revision 2.1a
and found that the loss of feedwater isolation signal for less than 3 days resulted in a
1E-7 (Green) outcome. The inspectors also determined that the cause of the finding has
a crosscutting aspect in the human performance area associated with decision making
because Wolf Creek failed to make a risk significant decision using a systematic
process. This issue was evaluated more than once and those evaluations sought to
justify bypassing the interlock rather than seek the full regulatory basis for the interlock.
- 20 - Enclosure
Enforcement. Wolf Creek Technical Specification, Table 3.3.2.1, function 8 includes
engineered safety features actuation system interlocks. Function 8.a, the P-4 interlock,
requires two trains to be operable in Modes 1, 2, and 3. Function 8.a does not provide a
required action for both trains of engineered safety features actuation system interlocks
inoperable. Wolf Creek Technical Specification 3.0.3 requires the plant to be in Mode 4
within 13 hours1.50463e-4 days <br />0.00361 hours <br />2.149471e-5 weeks <br />4.9465e-6 months <br /> if there is no required action specified for a limiting condition of operation
that cannot be met. Contrary to the above, from August 22 to August 23, 2009,
Wolf Creek failed to change modes from Mode 3 to Mode 4 when both trains of
engineered safety features actuation system interlock function 8.a, P-4, were inoperable
for greater than 13 hours1.50463e-4 days <br />0.00361 hours <br />2.149471e-5 weeks <br />4.9465e-6 months <br />. Specifically, from August 22 to 23, 2009, Wolf Creek failed to
change modes from Mode 3 to Mode 4 when both trains were removed from service for
approximately 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br />. Because this violation was determined to be of very low safety
significance and was placed in the corrective action program as Condition Report 19318,
this violation is being treated as a noncited violation in accordance with Section VI.A.1 of
the Enforcement Policy: NCV 05000482/2009004-04, Failure to Implement Engineered
Safety Features Actuation System Technical Specification Results in Missed Mode
Change.
1R17 Evaluations of Changes, Tests, or Experiments and Permanent Plant
Modifications (71111.17)
a. Inspection Scope
The inspectors reviewed the effectiveness of the licensees implementation of
evaluations performed in accordance with 10 CFR 50.59, Changes, Tests, and
Experiments, and changes, tests, experiments, or methodology changes that the
licensee determined did not require 10 CFR 50.59 evaluations. The inspection
procedure requires the review of 6 to 12 licensee evaluations required by 10 CFR 50.59,
12 to 25 changes, tests, or experiments that were screened out by the licensee and 5 to
15 permanent plant modifications.
The inspectors reviewed 9 evaluations required by 10 CFR 50.59. These included:
- 2006-001, Radiological Consequences of a Fuel Handling Accident, Revision 0
- 2008-0006, Wolf Creek Generating Station (WCGS) Simplified Head Assembly
(SHA) Drop Analysis, Revision 0
- 2008-0008, Use of Dedicated Operator for SI Pump B Room cooler
Replacement, Revision 0
- 2005-004, WCGS Rod Withdrawal at Power Event Safety Analysis, Revision 0
- 2008-001, Evaluations of Voids in the ECCS Suction Piping, Revision 0
- 2008-002, Evaluations of Voids in the ECCS Discharge Piping, Revision 0
- 2006-002, Power Operation, Revision 54
- 21 - Enclosure
- 2008-0003, Use of Dedicated Operator for SI Pump A Room Cooler
Replacement , Revision 0
- 2008-0004, MSFIS Controls Replacement , Revision 0
The inspectors reviewed 17 changes, tests, and experiments that were screened out by
licensee personnel. These included:
- CP 12820, Containment Room Cooler SGN01D, Revision 1
- CP 12876, Main Steam Atmospheric Relief Valve Aux (Pilot) Plug and Main Plug
Machining Dimensions, Revision 0
for the LTOP system, and the PTLR, Revision 0
- CP 13089, EF-138-HBC-30 Essential Service Water Pipe Pit Encapsulation,
Revision 1
- CP 11987, EKJ03A/B Replacement Heat Exchangers, Revision 6
- CP 12758, Coating Degradation and Isolated Pitting of Containment Incore
Instrumentation Sump Layer, Revision 3
- CP 12273, Shrinkage Effect at the Pressurizer Spray Nozzle on TBB03 Due to
Weld-Overlay, Revision 3
- CP 12489, SGK05A Tube Sheet and Channel Cover Degradation Evaluation,
Revision 0
- CP 12341, Region 19 Fuel Assembly and Core Component Configuration
Changes, Revision 0
- CP 12154, Relocate CVT Level transmitter BGLT0185, Revision 3
EMHV8803B, Revision 0
- CP 12639, 9 Volt Power Supply for SP067 & SP010 , Revision 0
- 22 - Enclosure
- TMP 08-022 , SI Accumulator C Boron Concentration Adjustment, Revision 0
The inspectors reviewed 7 permanent plant modifications. These included:
- CP 11987, EKJ03A/B Replacement Heat Exchangers, Revision 6
- CP 11379, Replacement for Obsolete Rad Monitoring Transducer, Revision 2
- CP 13089, EF-138-HBC-30" Essential Service Water Pipe Pit Encapsulation,
Revision 1
Revision 1
- CP 9488, Governor Replacement on Emergency Diesel Generators, Revision 7
- CP 11897, Transformer XNB02 Tap Change, Revision 2
The inspectors verified that when changes, tests, or experiments were made, that
evaluations were performed in accordance with 10 CFR 50.59 and that licensee
personnel had appropriately concluded that the change, test or experiment can be
accomplished without obtaining a license amendment. The inspectors also verified that
safety issues related to the changes, tests, or experiments were resolved. The
inspectors reviewed changes, tests, and experiments that licensee personnel
determined did not require evaluations and verified that the licensee personnels
conclusions were correct and consistent with 10 CFR 50.59. The inspectors also
verified that procedures, design, and licensing basis documentation used to support the
changes were accurate after the changes had been made and that preparers and
reviewers of the evaluations and screens were qualified and certified in accordance with
licensee procedures.
During the portion of the inspection dealing with modifications, the inspectors verified
that supporting design and license basis documentation had been updated accordingly
and was still consistent with the new design. The inspectors verified that procedures,
training plans and other design basis features had been adequately accounted for and
updated. Additional documents reviewed during this inspection are listed in the
attachment.
The inspectors verified that the licensee was identifying permanent plant modification
issues and problems related to 10 CFR 50.59 applicability determinations, screenings
and evaluations, and had entered them in the corrective action program. The inspectors
selected several samples to evaluate the appropriateness of the corrective actions
program. No program concerns were identified with corrective action documents
reviewed.
These activities constitute completion of one sample as defined in IP 71111.17-05
- 23 - Enclosure
b. Findings
No findings of significance were identified.
1R18 Plant Modifications (71111.18)
a. Inspection Scope
The inspectors reviewed the following temporary/permanent modifications to verify that
the safety functions of important safety systems were not degraded:
- Emergency diesel Generator B oil collection, August 13, 2009
- Heat tracing for the boric acid system, March 26, 2009
The inspectors reviewed the temporary modification and the associated safety
evaluation screening against the system design bases documentation, including the
USAR and the technical specifications, and verified that the modification did not
adversely affect the system operability/availability. The inspectors also verified that the
installation and restoration were consistent with the modification documents and that
configuration control was adequate. Additionally, the inspectors verified that the
temporary modification was identified on control room drawings, appropriate tags were
placed on the affected equipment, and licensee personnel evaluated the combined
effects on mitigating systems and the integrity of radiological barriers.
These activities constitute completion of two samples for temporary plant modifications
as defined in IP 71111.18-05.
b. Findings
Introduction. The inspectors identified a Green noncited violation of 10 CFR Part 50,
Appendix B, Criterion III, Design Control, for failing to translate the boric acid design
basis into time sensitive operator actions to ensure the core operating limits report
shutdown margin can be achieved with the boric acid flow path.
Description. On March 27, 2009, the inspectors walked down the safety injection pump
Room A and noted a temporary modification of heat tracing installed on boric acid piping.
The heat tracing was plugged into a nonsafety-related wall outlet for power. From the
boric acid tanks, the highly concentrated boric acid piping travels to the safety injection
pump Room A and then to the centrifugal charging pump suctions. The inspectors
reviewed the temporary modification documentation and found that Wolf Creek had
written Performance Improvement Request 2005-3461 in December 2005, stating that
this piping carried boric acid. Performance Improvement Request 2005-3461 identified
that, if the room coolers were started while lake temperature was low, the room
temperature may decrease below the solubility limit. It also identified that compensatory
actions may be needed. Corrective actions for heat tracing and instructions to operators
took approximately 3 years to implement, and stopped short of addressing boric acid
system operation when nonsafety power is lost to the heat tracing and the plant must be
taken to cold shutdown in accordance with technical specifications or plant conditions.
Achieving cold shutdown using only safety-related components is consistent with
Section 9.3 of the USAR. Control room operators had no procedural guidance to ensure
- 24 - Enclosure
that boration would be performed prior to the room and piping cooling to below the boric
acid precipitation temperature and blocking the piping. The core operating limits report
requires a cold shutdown margin of 1300 percent milirho (pcm). The inspectors found
that the procedural path to borating to cold shutdown conditions would likely take longer
than the time for the piping to cool to the boric acid precipitation temperature. Wolf
Creek performed an informal room heat loss calculation, but neglected forced cooling by
the room cooler, particularly with low lake temperature. The other boric acid source is
the refueling water storage tank which is not protected from external event such as
tornados. Therefore, the refueling water storage tank is not available in all safe
shutdown scenarios.
Wolf Creek also performed an informal simulator evaluation with licensed operators.
The scenario involved a loss of offsite power without the refueling water storage tank
available. The inspectors noted that the operators in the informal evaluation took less
time to arrive at the key boration steps in emergency procedures than the operators did
during an actual loss of offsite power event of August 19, 2009. The inspectors also
noted the August 19 event was less complicated than the simulator scenario, and the
simulator evaluation also did not involve emergency action level declarations or loss of
large portions of other equipment due to external events, such as a tornado. The
inspectors determined that these factors would add considerable time to that
demonstrated by the informal simulator evaluation. The inspectors concluded that the
licensee had failed to demonstrate that boration could be accomplished prior to boric
acid precipitation following a loss of nonsafety-related electrical power.
The inspectors also reviewed Procedure SYS BG-206, Boric Acid System Operation,
and found that the solubility limit for a 7680 parts per million boric acid solution is
63 degrees Fahrenheit. The inspectors found log entries from March 27, 2008, and
February 8, 2009, in which room temperature decreased to 67 and 58 degrees and
could have challenged the boric acid system by blocking the piping with precipitated
boron. However, the inspectors found that the refueling water storage tank was
operable and could have performed the reactivity control function in certain scenarios
that do not involve tornados or external events. Using these factors, inspectors
concluded that Wolf Creek had less time to accomplish more lengthy tasks in order to
perform boration to cold shutdown conditions.
The inspectors reviewed the corrective action history for heat tracing Temporary
Modification 07-012-BG. The inspectors reviewed Condition Report 2005-3461 and
found that it was continued under Condition Report 2007-2472. Condition
Report 2007-2472 created Corrective Action 4222 which was to plan and install heat
tracing under a temporary modification. The temporary modification installation work
order began on October 29, 2008. Condition Report 2007-2472 also had a corrective
action to issue guidance to nonlicensed operators taking temperature readings in the
safety injection pump Room A. These updated logs were implemented on December 19,
2008, and instructed operators that the boric acid piping may become inoperable due to
precipitation if room temperature dropped below 67 degrees Fahrenheit. There was no
guidance to operators in the control room regarding this time sensitive manual action.
Analysis. The failure to translate the design bases into procedures that ensure the
function of the safety-related boric acid system upon loss of nonsafety-related heat
- 25 - Enclosure
tracing is a performance deficiency. The inspectors determined that this finding was
more than minor because this issue aligned with Inspection Manual Chapter 0612,
Appendix E, example 2.f, because the pipe temperature was required to stay above the
boric acid solubility limit and the loss of the heat tracing and or room temperature
decrease will block the boric acid system. This issue was associated with the equipment
performance attribute of the mitigating systems cornerstone and affected the
cornerstone objective to ensure the availability, reliability, and capability of systems that
respond to initiating events. The inspectors evaluated the significance of this finding
using Phase 1 of Inspection Manual Chapter 0609, Appendix A, "Significance
Determination of Reactor Inspection Findings for At Power Situations," and determined
that the finding screened to phase 2 because the issue was a design or qualification
deficiency confirmed to result in loss of operability or functionality The inspectors
evaluated the significance of this finding using Phase 2 of Inspection Manual Chapter 0609, Risk Informed Inspection Notebook for Wolf Creek Generating Station, and
determined that the finding was of very low safety significance because loss of the boric
acid system in Table 3.9 for one year resulted in a 1E-7 CDF when giving recovery credit
for the refueling water storage tank. The inspectors determined that this finding has a
crosscutting aspect in the area of problem identification and resolution associated with
the corrective action program component because Wolf Creek did not take appropriate
corrective actions to resolve known deficiencies in the design and operation of the boric
acid system for approximately 4 years. The issue was re-evaluated in 2009 and failed to
correct the deficiencies identified in 2005 P.1.d].
Enforcement. Part 50 of Title 10 of the Code of Federal Regulations, Appendix B,
Criterion III, Design Control, requires, in part, that the design basis is correctly
translated into specifications, drawings and procedures. Achieving cold shutdown using
only safety-related components is consistent with Section 9.3 of the USAR. Contrary to
the above, since December 16, 2005, Wolf Creek has failed to ensure that the boric acid
system could perform its design function as specified in USAR, Section 9.3. Specifically,
Wolf Creek failed to ensure that time-sensitive operator actions to ensure the core
operating limits report specified shutdown margin can be achieved prior to boric acid
precipitates blocking the flow path. Because this violation is of very low safety
significance and has been entered into Wolf Creek's corrective action program as
condition report 20717, this violation is being treated as an noncited violation consistent
with Section VI.A.1 of the NRC Enforcement Policy: NCV 05000482/2009004-05, Use
of Nonsafety-Related Power to Ensure Operability of Safety-Related Boric Acid System.
1R19 Postmaintenance Testing (71111.19)
a. Inspection Scope
The inspectors reviewed the following postmaintenance activities to verify that
procedures and test activities were adequate to ensure system operability and functional
capability:
- Emergency Diesel Generator B run after compression fitting lube oil leak repaired
on August 17, 2009
- 26 - Enclosure
- Turbine-Driven auxiliary feedwater pump run after trip and throttle valve
maintenance on September 9, 2009
- Component cooling water train swaps after modification to valves on August 14,
2009
- Testing after repair to Emergency Diesel Generator A on December 5, 2008
- Replacement of Flow Transmitter BG FK-121 on August 28, 2009
- Limitorque and gearbox overhaul of essential service water Valve EF HV-31 on
August 31, 2009
- Essential service water Valve EF HV-42 after maintenance on August 12, 2009
- Safety Bus NB02 Channel 4 under-voltage relay power supply replacement on
March 24, 2009
The inspectors selected these activities based upon the structure, system, or
component's (SSC) ability to affect risk. The inspectors evaluated these activities for the
following:
- The effect of testing on the plant had been adequately addressed; testing was
adequate for the maintenance performed
- Acceptance criteria were clear and demonstrated operational readiness; test
instrumentation was appropriate
The inspectors evaluated the activities against the technical specifications, the USAR,
10 CFR Part 50 requirements, licensee procedures, and various NRC generic
communications to ensure that the test results adequately ensured that the equipment
met the licensing basis and design requirements. In addition, the inspectors reviewed
corrective action documents associated with postmaintenance tests to determine
whether the licensee was identifying problems and entering them in the corrective action
program and that the problems were being corrected commensurate with their
importance to safety. Specific documents reviewed during this inspection are listed in
the attachment.
These activities constitute completion of eight postmaintenance testing inspection
samples as defined in IP 71111.19-05.
b. Findings
Introduction. The inspectors identified a Green noncited violation of Technical
Specification 3.8.1.B.4 in which the licensee removed equipment from service that was
required by technical specifications and the NRC safety evaluation.
Description. On March 24, 2009, the licensee entered Technical Specification 3.8.1,
Required Action B.4.2.2. This action allowed an emergency diesel generator to be
inoperable for up to 7 days. On March 24, 2009, at 4:20 p.m., the inspectors noted that
- 27 - Enclosure
Wolf Creek performed Procedure STS IC-208B, 4kV Loss of Voltage and Degraded
Voltage TADOT NB02 Bus - Separation Group 4, Revision 2A, to determine the as-
found conditions of the Channel 4 under voltage power supply. Operators entered
Technical Specification 3.3.5, Condition A.1 and exited 19 minutes later. The power
supply voltage ripple passed Procedure STS IC-208B, but Wolf Creek elected to replace
it. Again on March 24, 2009, at 4:54 p.m., Wolf Creek entered Technical Specification 3.3.5, Condition A.1, to replace the subject Channel 4 power supply. Condition A.1
required the out-of-service channel to be placed in trip within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. Wolf Creek exited
Technical Specification 3.3.5 at 9:09 p.m., on March 24. The removal of Channel 4 from
service resulted in a higher probability of loss of power to the safety bus because the
coincidence logic changed from two out of four to one out of three. The inspectors found
that this logic was an input to the NB02 normal offsite power feeder breaker described in
the offsite power surveillance procedure, STS NB-005, Breaker Alignment Verification,
Revision 18.
The 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. The inspectors also reviewed the NRC
Safety Evaluation Report (SER) for the 7-day emergency diesel generator allowed
outage time (Technical Specification 3.8.1.B.4.2.2) and found that Section 4.6.c, states:
The offsite power supply [emphasis added] 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 emergency diesel generator completion time are removed from service, which
includes auxiliary feedwater, component cooling water, essential service water and their
support systems. The support equipment protections are also mirrored in Section 4.0 of
the NRC safety evaluation for Amendment 163. However, Wolf Creek removed one
channel of under voltage protection for offsite power to Bus NB02 (Train B) which is a
support system for the above equipment. The inspectors found that
Procedure STS IC-208B permits the testing of degraded voltage relays while the diesel
is out of service. These relays control the opening logic for the normal offsite power feed
to the safety bus NB02. 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. Normally
the safety bus NB02 cabinets are protected equipment (no work allowed) but because
this work was planned in advance for the diesel outage, the work was permitted. In
consultation with the Office of Nuclear Reactor Regulation, the inspectors concluded that
Procedure STS IC-208B and power supply replacement was inappropriate during the
7-day diesel outages because it increased the probability of the loss of offsite power to
safety equipment that could not be powered by the diesel. 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. The inspectors found that Wolf Creek assessed risk under
10 CFR 50.65 a(4) for this evolution, resulting in elevated risk within the Green band
during the 7-day diesel outage. The inspectors also found that Wolf Creek appropriately
- 28 - Enclosure
protected component cooling water, emergency service water, instrument busses, dc
busses, emergency core cooling, the Train A diesel, and control room ventilation.
The inspectors reviewed corrective actions from NCV 05000482/2008002-02 previously
identified by inspectors when Wolf Creek made one of the offsite power sources
inoperable during a 7-day diesel outage. The licensee reviewed
Procedure STS IC-208B but did not revise it because the load shedder and emergency
load sequencer procedure tests one channel at a time. No other expanded explanation
was articulated in Condition Report 2008-0489. Condition Report 15727 was initiated for
the March 24, 2009, maintenance, and the issue has since been corrected by Wolf
Creek.
Analysis. The inspectors determined that the failure to implement requirements of
Technical Specification 3.8.1 and the associated NRC safety evaluation 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 finding was more than minor because it is associated with the
equipment performance attribute for the Mitigating Systems Cornerstone and 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 finding was determined to be of very low safety significance because the
issue did not result in the Train B offsite power being inoperable for greater than
24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and did not involve external events such as flooding. Additionally, the cause of
the finding has a problem identification and resolution crosscutting aspect in the area
associated with the corrective action program. Specifically, Wolf Creek did an extent of
condition review in response to a previous violation which included
Procedure STS IC-208B, but still failed to prohibit performance of Procedure STS IC-
208B during 7-day diesel outages P.1(c).
Enforcement. Technical Specification 3.8.1, Required Action 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 are met. The NRC SER for License Amendment 163
requires that the offsite power supply and switchyard conditions be conducive to an
extended 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. Contrary to the above, on March 24, 2009,
Wolf Creek performed elective maintenance which challenged offsite power availability
while emergency diesel generator B was in the 7-day extended completion time.
Specifically the licensee performed maintenance on the safety bus NB02 degraded and
undervoltage voltage relay Channel 4 power supply while the emergency diesel
generator Train B was in an extended outage. Because the finding is of very low safety
significance and has been entered into the corrective action program as Condition
Report 15727, this violation is being treated as a noncited violation, consistent with
Section VI.A of the NRC Enforcement Policy: NCV 05000482/2009004-06, Performing
Prohibited Elective Maintenance on Safety Bus NB02 Channel 4 during Emergency
Diesel Generator Maintenance.
- 29 - Enclosure
1R20 Refueling and Other Outage Activities (71111.20)
a. Inspection Scope
The inspectors reviewed the outage safety plan and contingency plans for the Wolf
Creek outage conducted from August 19 to August 24, 2009, to confirm that licensee
personnel had appropriately considered risk, industry experience, and previous site-
specific problems in developing and implementing a plan that assured maintenance of
defense in depth. During the forced outage, the inspectors observed portions of the
shutdown and cooldown processes and monitored licensee controls over the outage
activities listed below.
- Configuration management, including maintenance of defense indepth, is
commensurate with the outage safety plan for key safety functions and
compliance with the applicable technical specifications when taking equipment
out of service.
- Clearance activities, including confirmation that tags were properly hung and
equipment appropriately configured to safely support the work or testing.
- Status and configuration of electrical systems to ensure that technical
specifications and outage safety-plan requirements were met, and controls over
switchyard activities.
- Monitoring of decay heat removal processes, systems, and components.
- Controls over activities that could affect reactivity.
- Startup and ascension to full power operation, tracking of startup prerequisites,
walkdown of the drywell (primary containment) to verify that debris had not been
left which could block emergency core cooling system suction strainers, and
reactor physics testing.
- Licensee identification and resolution of problems related to the August 19, 2009,
forced outage activities.
Specific documents reviewed during this inspection are listed in the attachment.
These activities constitute completion of one refueling outage and other outage
inspection sample as defined in IP 71111.20-05.
b. Findings
No findings of significance were identified.
- 30 - Enclosure
1R22 Surveillance Testing (71111.22)
a. Inspection Scope
The inspectors reviewed the USAR, procedure requirements, and technical
specifications to ensure that the four surveillance activities listed below demonstrated
that the SSCs tested were capable of performing their intended safety functions. The
inspectors either witnessed or reviewed test data to verify that the significant
surveillance test attributes were adequate to address the following:
- Preconditioning
- Evaluation of testing impact on the plant
- Acceptance criteria
- Test equipment
- Procedures
- Jumper/lifted lead controls
- Test data
- Testing frequency and method demonstrated technical specification operability
- Test equipment removal
- Restoration of plant systems
- Fulfillment of ASME code requirements
- Updating of performance indicator data
- Engineering evaluations, root causes, and bases for returning tested systems,
structures, and components not meeting the test acceptance criteria were correct
- Reference setting data
- Annunciators and alarms setpoints
The inspectors also verified that licensee personnel identified and implemented any
needed corrective actions associated with the surveillance testing.
- 4kV loss of voltage and degraded voltage TADOT NB02 bus, July 14, 2009
- Essential service water Pump A inservice test, August 13, 2009
- End of life moderator temperature coefficient measurement, August 28, 2009
- 31 - Enclosure
- August 12, 2009, missed surveillance for over power deltaT and over
temperature deltaT
Specific documents reviewed during this inspection are listed in the attachment.
These activities constitute completion of four surveillance testing inspection samples as
defined in IP 71111.22-05.
b. Findings
No findings of significance were identified.
4. OTHER ACTIVITIES
4OA1 Performance Indicator Verification (71151)
.1 Data Submission Issue
a. Inspection Scope
The inspectors performed a review of the data submitted by the licensee for the 2nd
Quarter 2009 performance indicators for any obvious inconsistencies prior to its public
release in accordance with Inspection Manual Chapter 0608, Performance Indicator
Program.
This review was performed as part of the inspectors normal plant status activities and,
as such, did not constitute a separate inspection sample.
b. Findings
No findings of significance were identified.
.2 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 1st quarter 2008 through the
2nd quarter 2009. To determine the accuracy of the performance indicator data reported
during those periods, performance indicator definitions and guidance contained in
Nuclear Energy Institute Document 99-02, Regulatory Assessment Performance
Indicator Guideline, Revision 5, was used. The inspectors reviewed the licensees
operator narrative logs, issue reports, event reports, and NRC integrated inspection
reports for the period of January 1, 2008, through June 30, 2009, to validate the
accuracy of the submittals. The inspectors also reviewed the licensees issue report
database to determine if any problems had been identified with the performance
indicator data collected or transmitted for this indicator and none were identified.
Specific documents reviewed are described in the attachment to this report.
These activities constitute completion of one unplanned scrams with complications
sample as defined by IP 71151-05.
- 32 - Enclosure
b. Findings
Wolf Creek will submit a Frequently Asked Question to determine if the April 19, 2009,
unplanned scram should also be counted as a scram with complications.
.3 Safety System Functional Failures
a. Inspection Scope
The inspectors sampled licensee submittals for the safety system functional failures
performance indicator for the period from the 1st quarter 2008 through the 2nd quarter
2009. To determine the accuracy of the performance indicator data reported during
those periods, performance indicator definitions and guidance contained in NEI
Document 99-02, Regulatory Assessment Performance Indicator Guideline, Revision 5,
and NUREG-1022, Event Reporting Guidelines 10 CFR 50.72 and 50.73" definitions
and guidance were used. The inspectors reviewed the licensees operator narrative
logs, operability assessments, maintenance rule records, maintenance work orders,
issue reports, event reports and NRC Integrated Inspection reports for the period of
January 1, 2008, through June 30, 2009, 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 three were identified. Specific documents reviewed are
described in the attachment to this report.
These activities constitute completion of one safety system functional failures sample as
defined by IP 71151-05.
b. Findings
The inspectors identified one violation of 10 CFR 50.73(a)(2)(v) with three examples.
This section of the rule is the NEI 99-02 definition of a safety system functional failure.
The enforcement aspects of this violation are discussed in Section 4OA3 of this report.
4OA2 Identification and Resolution of Problems (71152)
Cornerstones: Initiating Events, Mitigating Systems, Barrier Integrity, Emergency
Preparedness, Public Radiation Safety, Occupational Radiation Safety, and Physical
Protection
.1 Routine Review of Identification and Resolution of Problems
a. Inspection Scope
As part of the various baseline inspection procedures discussed in previous sections of
this report, the inspectors routinely reviewed issues during baseline inspection activities
and plant status reviews to verify that they were being entered into the licensees
corrective action program at an appropriate threshold, that adequate attention was being
given to timely corrective actions, and that adverse trends were identified and
addressed. The inspectors reviewed attributes that included: the complete and
accurate identification of the problem; the timely correction, commensurate with the
- 33 - Enclosure
safety significance; the evaluation and disposition of performance issues, generic
implications, common causes, contributing factors, root causes, extent of condition
reviews, and previous occurrences reviews; and the classification, prioritization, focus,
and timeliness of corrective actions. Minor issues entered into the licensees corrective
action program because of the inspectors observations are included in the attached list
of documents reviewed.
These 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 followup, the inspectors performed a daily screening of
items entered into the licensees corrective action program. The inspectors
accomplished this through review of the stations daily corrective action documents.
The inspectors performed these daily reviews as part of their daily plant status
monitoring activities and, as such, did not constitute any separate inspection samples.
b. Findings
No findings of significance were identified.
.3 Selected Issue Followup Inspection
a. Inspection Scope
During a review of items entered in the licensees corrective action program, the
inspectors recognized a corrective action item documenting an experimental test to
resolve the condition of the reactor coolant pump thermal barriers identified in cited
violation: NOV 05000482/2009002-07.
These activities constitute completion of one in depth problem identification and
resolution sample as defined in IP 71152-05.
b. Findings
No findings of significance were identified.
- 34 - Enclosure
4OA3 Event Follow-up (71153)
.1 Loss of Offsite Power and Reactor Trip on August 19, 2009
a. Inspection Scope
On August 19, 2009, inspectors responded to a reactor trip and a loss of offsite power
when the 345 kV La Cygne line was struck by lightning. The inspectors verified that the
emergency diesel generators started and supplied loads. The inspectors monitored
control room activities and equipment until normal offsite power feeds were re-aligned to
the safety busses. The inspectors walked down portions of the plant to ensure safety
systems were functioning.
These activities constitute completion of one event response sample as defined in
IP 71153-05.
b. Findings
No findings of significance were identified. This event was reviewed in detail by an NRC
special inspection team. The results of the special inspection will be documented in
NRC Inspection Report 2009-007.
.2 Failure to Report Conditions that Could Have Prevented Fulfillment of a Safety Function
a. Inspection Scope
The inspectors implemented IP 71151 consistent with Section 4OA1 of this report. The
inspectors also utilized IP 71153 to review licensee event reports. The findings are
documented below in accordance with Inspection Manual Chapter 0612.
b. Findings
Introduction. The inspectors identified a Severity Level IV noncited violation of 10 CFR
50.73, with three examples in which the licensee failed to submit licensee event reports
within 60 days following discovery of events or conditions meeting the reportability
criteria.
Description. First, on April 10, 2008, the licensee submitted LER 2008-002 under
10 CFR 50.73(a)(2)(i)(B) which is operation prohibited by technical specifications. For
11 hours1.273148e-4 days <br />0.00306 hours <br />1.818783e-5 weeks <br />4.1855e-6 months <br /> from February 13-14, 2008, Wolf Creek did not have an operable emergency
core cooling system because no high head charging pumps were operable. Wolf Creek
was in Technical Specification 3.0.3 during this time. Wolf Creek received enforcement
discretion to remain at power. Charging Pump B was required to be declared inoperable
because emergency diesel generator B was inoperable, and charging Pump A was
inoperable because it did not have an operable room cooler. On June 25, 2009, the
inspectors identified that Wolf Creek failed to report this event as a safety system
functional failure under 10 CFR 50.73(a)(2)(v) for the emergency core cooling system
being inoperable. The inspectors discussed this with Wolf Creek and Condition
Report 00018156 was initiated. On July 30, 2009, the licensee completed the evaluation
- 35 - Enclosure
of this condition report and concluded that the loss of high head charging was not
reportable, however no evaluation demonstrated operability of the charging pumps.
The inspectors reviewed this issue under the safety system functional failures
performance indicator. NEI 99-02, Regulatory Assessment Performance Indicator
Guideline, Revision 5, defines a safety-system functional failure as those events meeting
10 CFR 50.73(a)(2)(v) and requires evaluation of conditions reported under other
paragraphs of 50.73 for safety-system functional failures. Wolf Creek did not perform a
review. Wolf Creek subsequently drafted a position paper which relied on the
statements made in the Letter WO 08-0006, Request for Notice of Enforcement
Discretion from Technical Specification 3.8.1, AC Sources - Operating, which
contained an attachment that provided information documenting Wolf Creeks verbal
request for the Enforcement Discretion. The attachment contained the risk mitigation
manual actions for not shutting down the unit, a discussion of the calculated incremental
core damage probability used to justify enforcement discretion, and a qualitative
statement regarding the adjacent pumps room coolers. Wolf Creek also stated that it
considered the centrifugal charging pump to be functional. The manual actions did not
involve the failed room cooler. Wolf Creek also cited LER 2008-002-00 which contained
the same discussion of the risk assessment, the functionality of the charging Pump A,
and the adjacent pumps room coolers. The inspectors did not find an evaluation
demonstrating the operability of charging Pump A or B and hence the emergency core
cooling system.
The inspectors consulted NUREG 1022, Event Reporting Guidelines 10 CFR 50.72 and
50.73, Revision 2. NUREG 1022 Section 3.2.7, reportability under 50.73(a)(2)(v),
states that operability under Generic Letter 91-18 is the correct standard to apply.
Generic Letter 91-18 has been superseded by Regulatory Issue Summary 2005-20
which does not permit the use of risk assessment to justify operability. The inspectors
found that Wolf Creek was incorrect in concluding that the application of functional under
the risk assessment was equivalent to the words of safety function under
50.73(a)(2)(v). Another position paper drafted by Wolf Creek stated that centrifugal
charging Pump B was operable although it was not supported by an operable
emergency diesel generator. The inspectors disagreed with this application of the
definition of the technical specification of operability and this application of Technical Specifications 3.8.1, 3.0.2, and 3.0.6 which require equipment to be supported by
emergency power to perform the safety function. The inspectors consulted with NRR,
who agreed with the inspectors use of the rule and NUREG 1022. The issue was again
placed into the corrective action program as Condition Report 19914.
In the second example, Wolf Creek filed LER 2008-004-00 on June 6, 2008. LER 2008
004-00 was filed under 50.73(a)(2)(iv)(A) for an event that caused automatic start of an
emergency diesel during a loss of offsite power on April 16, 2008. No report was made
under 50.73(a)(2)(v) for an event or condition that could have prevented a safety
function due to the loss of offsite power. Inspectors reviewed NUREG 1022,
Section 3.2.7 and found that:
"Both offsite electrical power (transmission lines) and onsite emergency power
(usually diesel generators) are considered to be separate functions by GDC 17. If
either offsite power or onsite emergency power is unavailable to the plant, it is
- 36 - Enclosure
reportable regardless of whether the other system is available. GDC 17 defines
the safety function of each system as providing sufficient capacity and capability,
etc., assuming that the other system is not available. Loss of offsite power should
be determined at the essential switchgear busses."
This missed licensee event report is specifically captured in Condition Report 19371.
Wolf Creek indicated that it plans to update LER 2008-004-00 or make a second
licensee event report.
Third, on April 10, 2008, Wolf Creek filed Event Notification Report 44131 per 10 CFR 50.72(b)(3)(ii)(B) based on a possible trip of all four containment coolers. The
containment coolers have thermal overload protection such that if a cooler trips in fast
speed during normal power operation, that cooler will not restart in slow speed for an
accident. Wolf Creek evaluated this concern and issued Event Notification 44131. Wolf
Creek later retracted the Event Notification stating: "Further analysis of the main steam
line break, if this concern had existed, showed that the calculated post-accident pressure
and temperature peak values would not exceed the peak accident values in the USAR.
Therefore, an unanalyzed condition did not exist and Wolf Creek is retracting the
50.72(b)(3)(ii)(B) notification."
The inspectors found that Wolf Creek did not analyze the current draw for the motors
prior to receipt of a safety injection signal. Wolf Creek assumed that the coolers would
not restart and relied on containment, but this is still the loss of a safety function to
remove heat from containment. Wolf Creek found that without the coolers, containment
pressure exceeds the Analysis of Record but not the design pressure in the USAR.
Inspectors found that this was not an appropriate method to consider the coolers heat
removal safety function met. At the end of the report period, Wolf Creek did not have an
analysis for the containment cooler motors to determine if they would have tripped prior
to receiving an accident signal. Wolf Creeks condition report and reportability
evaluation has been open since April 11, 2008. No licensee event report has been
submitted. The inspectors found insufficient evidence to show that the containment
coolers could accomplish their safety function and that this should have been reported
under 10 CFR 50.73(a)(2)(v). This issue is captured in Condition Report 15318.
Analysis. The failure to submit a licensee event report was a performance deficiency.
The inspectors reviewed this issue in accordance with Inspection Manual Chapter 0612
and the NRC Enforcement Manual. Through this review, the inspectors determined that
traditional enforcement was applicable to this issue because the NRC's regulatory ability
was affected. Specifically, the NRC relies on the licensee to identify and report
conditions or events meeting the criteria specified in regulations in order to perform its
regulatory function, and when this is not done, the regulatory function is impacted. The
inspectors determined that this finding was not suitable for evaluation using the
significance determination process, and as such, was evaluated in accordance with the
NRC Enforcement Policy. The finding was reviewed by NRC management, and because
the violation was determined to be of very low safety significance, was not repetitive or
willful, and was entered into the corrective action program, this violation is being treated
as a Severity Level IV noncited violation consistent with the NRC Enforcement Policy.
This finding was determined to have a crosscutting aspect in the area of problem
identification and resolution associated with the corrective action program in that the
- 37 - Enclosure
licensee failed to appropriately and thoroughly evaluate for reportability aspects all
factors and time frames associated with the inoperability of the emergency core cooling
system, the offsite power system, and the containment heat removal system P.1(c)
(4OA3)
Enforcement. Title 10 CFR 50.73(a)(1) requires, in part, that licensees shall submit a
licensee event report for any event of the type described in this paragraph within 60 days
after the discovery of the event. Title 10 CFR 50.73(a)(2)(v) requires, in part, that events
or conditions that could have prevented the fulfillment of the safety function of structures
or systems that are needed to shutdown the reactor and maintain it in a safe shutdown
condition, remove residual heat, control the release of radioactive material, or mitigate
the consequences of an accident. Contrary to the above, in 2008, Wolf Creek failed to
submit a licensee event report within 60 days for three separate events that could have
prevented the fulfillment of the safety function of structures or systems that are needed
to shutdown the reactor and maintain it in a safe shutdown condition, remove residual
heat, control the release of radioactive material, or mitigate the consequences of an
accident. Specifically, emergency core cooling, offsite power, and containment cooling
could have been or were actually lost on February 13-14, 2008, April 16, 2008, and
April 10, 2008, respectively, and Wolf Creek did not submit an LER within 60 days. Wolf
Creek did not have sufficient analyses to demonstrate that these three events were not
reportable. In accordance with the NRC's Enforcement Policy, the finding was reviewed
by NRC management and because the violation was of very low safety significance, was
not repetitive or willful, and was entered into the corrective action program, this violation
is being treated as a Severity Level IV noncited violation, consistent with the NRC
Enforcement Policy: NCV 05000482/2009004-07, Failure to Report Conditions that
Could Have Prevented Fulfillment of a Safety Function.
4OA5 Other Activities
.1 Quarterly Resident Inspector Observations of Security Personnel and Activities
a. Inspection Scope
During the inspection period, the inspectors performed observations of security force
personnel and activities to ensure that the activities were consistent with Wolf Creek
security procedures and regulatory requirements relating to nuclear plant security.
These observations took place during both normal and off-normal plant working hours.
These quarterly resident inspector observations of security force personnel and activities
did not constitute any additional inspection samples. Rather, they were considered an
integral part of the inspectors normal plant status review and inspection activities.
b. Findings
No findings of significance were identified.
- 38 - Enclosure
.2 INPO Training Program Accreditation
a. Inspection Scope
The NRC reviewed the concerns raised by the Accreditation Board. The senior resident
inspector read WANO (INPO) accreditation report and discussed the issues with the
licensee and NRR and determined that there were not any safety significant training
deficiencies. The NRC determined that compliance with the regulations is not affected
and that the probationary status is not safety significant. No further NRC action is
required under Inspection Procedure 41500.
b. Findings
No findings of significance were identified.
.3 (Closed) Unresolved Item 05000482/2008010-03: Changes to the Approved Fire
Protection Program May Not Meet NRC Acceptance Criteria
Introduction. The inspectors identified a Severity Level IV noncited violation of License
Condition 2.C.(5), Fire Protection, for making changes to the approved fire protection
program without the required prior Commission approval. Specifically, the licensee
made a change to the USAR that allowed the licensee to violate the requirements of
10 CFR Part 50, Appendix R, Section III.L.
Description. During the 2005 triennial fire protection inspection, the team identified an
apparent violation concerning the failure to ensure that the reactor coolant system would
not lose subcooling during an alternative shutdown scenario if a fire caused both
pressurizer power operated relief valves to spuriously open. This issue was
documented as Apparent Violation 05000482/2005008-02, Failure to Maintain Reactor
Coolant System Subcooling During the Alternative Shutdown.
After the 2005 inspection, the licensee made significant changes to the alternative
shutdown methodology implemented by Procedure OFN RP-017, Control Room
Evacuation. The licensee also developed Report E-1F9915, Design Basis Document
for OFN RP-017, Control Room Evacuation, Revision 0, and Evaluation SA-08-006,
RETRAN-3D Post-Fire Safe Shutdown (PFSSD) Consequence Evaluation for a
Postulated Control Room Fire, Revision 0, to demonstrate the adequacy of the revised
alternative shutdown procedure. These evaluations predicted that a fire in the control
room which led to control room abandonment and caused a single pressurizer power
operated relief valve to spuriously open could cause a steam bubble to void
approximately 40 percent of the reactor vessel head.
In response to these evaluations, the licensee modified the fire protection program to
allow voiding in the core. Specifically, the licensee modified Table 9.5E-1 of the USAR
to include the following paragraph:
Analysis demonstrates that the performance goals of III.L.2 are satisfied.
The performance criteria of III.L.1 are also satisfied, with the exception of
maintaining reactor process variables within those predicted for a loss of
- 39 - Enclosure
normal ac power. This is acceptable, as long as a control room fire will
not result in the plant reaching an unrecoverable condition, which could
lead to core damage.
During the 2008 triennial fire protection inspection, the team identified an unresolved
item related to this change to the fire protection program. The team was concerned that
the licensee changed the fire protection program in a manner that could adversely affect
the ability to achieve and maintain safe shutdown in the event of a fire without prior NRC
approval. This concern was documented as Unresolved Item 05000482/2008010-03,
Changes to the Approved Fire Protection Program May Not Meet NRC Acceptance
Criteria.
The licensee stated that their original approved fire protection program was based on the
plant not reaching an unrecoverable condition during an alternative shutdown, citing
Letter SLNRC 84-109, dated August 23, 1984.
The staff reviewed the approved fire protection program, as specified by License
Condition 2.C.(5), and concluded the licensee was required to meet the technical
requirements of 10 CFR Part 50, Appendix R, Section III.L. As noted in License
Condition 2.C.(5), the approved fire protection program is described by the USAR
through Revision 17, the Wolf Creek site addendum through Revision 15, and the SER
through Supplement 5. In the Wolf Creek SER (NUREG-0881), Supplement 3, the staff
concluded that the alternative shutdown capability for the control room met the
requirement of 10 CFR Part 50, Appendix R, Section III.L, and was, therefore,
acceptable.
The staff also concluded that the standard not reaching an unrecoverable condition
was not part of the approved fire protection program, nor was the phrase no
unrecoverable condition used in the context of alternative shutdown in any of the three
documents specified in License Condition 2.C.(5). Further, the staff noted that the
licensee did not identify this as a deviation from the requirements of 10 CFR Part 50,
Appendix R,Section III.L.1, nor did the staff acknowledge any such deviation in their
approval of the alternative shutdown approach in the safety evaluation reports.
Section III.L of 10 CFR Part 50, Appendix R specifies:
During the postfire shutdown, the reactor coolant system process
variables shall be maintained within those predicted for a loss of normal
ac power, and the fission product boundary integrity shall not be affected;
i.e., there shall be no fuel clad damage, rupture of any primary coolant
boundary, of rupture of the containment boundary.
The team noted the plant response to a loss of normal ac power was described in the
USAR, Chapter 15, Section 15.2.6. The USAR indicated that the plant would maintain
reactor coolant system subcooling and no void formation would occur in the reactor
vessel head during a loss of normal ac power. Therefore, a change to the fire protection
program that allowed voiding in the reactor vessel head during an alternative shutdown
would involve a failure to meet the requirements of 10 CFR Part 50, Appendix R,
Section III.L.
- 40 - Enclosure
The staff reviewed the licensees program change and concluded that this change
exceeded the licensees ability to make changes without prior staff approval, as provided
in License Condition 2.C.(5). Specifically, the staff considers a change that allows the
licensee to violate a requirement to be a change that adversely affects the ability to
achieve and maintain safe shutdown in the event of a fire.
Analysis. Changing the approved fire protection program such that the reactor coolant
subcooling process variables would remain within those predicted for a loss of normal ac
power without prior Commission approval was a performance deficiency. The team
assessed this performance deficiency using traditional enforcement since it had the
potential for impacting the NRCs ability to perform its regulatory function. The team
determined this performance deficiency was more than minor since the change required
prior staff review and approval prior to implementation and it did not receive the required
approval.
A senior reactor analyst performed a Phase 3 evaluation to determine the risk
significance of this finding since the performance deficiency involved a control room fire
that led to control room abandonment. The analyst performed a bounding evaluation to
determine an upper limit for the change in core damage frequency.
The analyst assigned a generic fire ignition frequency for the control room (FIFCR), which
was slightly higher than the value in Calculation AN-95-029, Control Room Fire
Analysis, Revision 1. The analyst multiplied the fire ignition frequency by a severity
factor (SF) and a nonsuppression probability indicating that operators failed to extinguish
the fire within 20 minutes assuming a 2 minute detection that required a control room
evacuation (NPCRE). The resulting control room evacuation frequency (FEVAC) was:
FEVAC = FIFCR * SF * NPCRE
= 1.09E-2/year * 0.1 * 1.30E-2
= 1.42E-5/year
The control room has a total of 103 cabinets. The analyst determined that a single fire in
five of these cabinets could lead to the spurious opening of a pressurizer power-
operated relief valve. Therefore, a bounding change in core damage frequency for a
control room fire that leads to evacuation and the spurious opening of a pressurizer
power-operated relief valve (FEVAC+PORV) was determined to be:
FEVAC+PORV = FEVAC * 5 / 103
= 1.42E-5/year * 5 / 103
= 6.88E-7/year
- 41 - Enclosure
This frequency was considered to be bounding since it assumed:
1) A fire in the applicable cabinets would create a short that caused the
pressurizer power-operated relief valve to spuriously open,
2) The conditional core damage probability given a control room fire with
evacuation and the spurious opening of a power-operated relief valve was set
equal to one, and
3) The performance deficiency accounted for the entire change in core damage
frequency (i.e., the baseline core damage frequency for this event was zero).
Since this bounding frequency was less than 1E-6/year, the analyst determined this
performance deficiency to have very low risk significance.
This performance deficiency was analogous to Example D.5 in the Enforcement Policy,
Supplement 1. Since, the performance deficiency was evaluated as having very low
safety significance, the team determined that a Severity Level IV violation was
appropriate.
This finding had a crosscutting aspect in the area of human performance associated with
resources because the licensee failed to maintain long term plant safety by maintaining
design margins. Specifically, the licensees choice to allow reactor vessel head voiding
during an alternative shutdown in lieu of restoring the plant to compliance with the
requirements of 10 CFR Part 50, Appendix R, Section III.L constituted a reduction in
safety margin (H.2(a)).
Enforcement. License Condition 2.C.(5), Fire Protection, states, in part:
a) The operating corporation shall maintain in effect all provisions of the
approved fire protection program as described in the SNUPPS Final Safety
Analysis Report for the facility through Revision 17, the Wolf Creek site
addendum through Revision 15, and as approved in the SER through
Supplement 5, subject to provisions b & c below.
b) The licensee may make changes to the approved fire protection program
without prior approval of the Commission only if those changes would not
adversely affect the ability to achieve and maintain safe shutdown in the
event of a fire.
The SER, Section 9.5.1.7 states, in part:
The staff will condition the operating license to require the applicant to meet the
technical requirements of Appendix R to 10 CFR Part 50, or provide equivalent
protection.
- 42 - Enclosure
The SER, Supplement 3, Section 9.5.1.5 states:
Based on our review, the staff concludes that the alternative shutdown capability
for the control room meets the requirements of Appendix R,Section III.L, and is
therefore acceptable.
Section III.L of 10 CFR Part 50, Appendix R, specifies:
During the postfire shutdown, the reactor coolant system process variables shall
be maintained within those predicted for a loss of normal ac power, and the
fission product boundary integrity shall not be affected; i.e., there shall be no fuel
clad damage, rupture of any primary coolant boundary, or rupture of the
containment boundary.
The plant response to a loss of normal ac power was described in the USAR,
Chapter 15, Section 15.2.6. The USAR indicated that the plant would maintain reactor
coolant system subcooling and no void formation would occur in the reactor vessel head
during a loss of normal ac power.
Contrary to the above, on September 25, 2008, the licensee made a change to the
approved fire protection program that adversely affected the ability to achieve and
maintain safe shutdown in the event of a fire without prior approval of the Commission.
Specifically, the licensee made a change to Table 9.5E-1 of the USAR that allowed
reactor coolant system process variables to exceed those predicted for a loss of normal
ac power during an alternative shutdown. This change adversely affected the ability to
achieve and maintain safe shutdown in the event of a fire since it allowed the licensee to
violate a requirement without an approved deviation.
The licensee entered this issue into their corrective action program as Performance
Improvement Request 2008-004869. Because this violation was of very low safety
significance and it was entered into the licensees corrective action program, this
violation is being treated as a noncited violation, consistent with the NRC Enforcement
Policy: NCV 05000482/2009004-08, Changes to the Approved Fire Protection Program
Without Prior Staff Approval.
.4 (Closed) Apparent Violation 05000482/2005008-02: Failure to Maintain Reactor Coolant
System Subcooling During the Alternative Shutdown
The issue documented by this apparent violation is enveloped by Unresolved
Item 05000482/2008010-03, Changes to the Approved Fire Protection Program May
Not Meet NRC Acceptance Criteria and discussed in Section 4OA5.1. This apparent
violation is closed.
4OA6 Meetings
Exit Meeting Summary
On July 30, 2009, the inspectors presented the inspection results to Mr. S. A. Henry,
Manager, Plant Operations, and other members of the licensee staff. The inspectors
- 43 - Enclosure
stated that they had reviewed proprietary information during the inspection, and verified
that all material had been returned to the licensee or destroyed. The licensee
acknowledged the inspection results as presented.
The inspector briefed Robert Evenson of the results of the annual licensed operator
requalification program inspection on August 5, 2009. The licensee representative
acknowledged the findings presented. The inspectors asked the licensee whether any
materials examined during the inspection should be considered proprietary. No
proprietary information was identified.
On September 28, 2009, the inspectors conducted a telephonic exit meeting and
presented the results of the staff review of fire protection program changes to
Mr. J. Suter, Fire Protection Supervisor, and other members of the licensee staff. The
licensee acknowledged the issues presented. The inspectors asked the licensee
whether any of the material examined during the inspection should be considered
proprietary. No proprietary information was identified.
On October 14, 2009, the resident inspectors presented the inspection results of the
resident inspections to Mr. Matt Sunseri, Vice President Oversight, and other members
of the licensee's management staff. The licensee acknowledged the findings presented.
The inspectors noted that while proprietary information was reviewed, none would be
included in this report and that the materials were returned to the licensee.
- 44 - Enclosure
SUPPLEMENTAL INFORMATION
KEY POINTS OF CONTACT
Licensee Personnel
R. A. Muench, President and Chief Executive Officer
M. Sunseri, Vice President Operations and Plant Manager
S. E. Hedges, Vice President Oversight
G. J. Pendergrass, Manager Engineering
T. East, Manager, Emergency Planning
P. Bedgood, Superintendent, Chemistry/Radiation Protection
LIST OF ITEMS OPENED AND CLOSED
Opened and Closed
05000482/2009004-01 NCV Failure to Log Foreign Material in Spent Fuel Pool After
Extent of Condition Evaluation (Section 1R05)05000482/2009004-02 NCV Inability to Perform Manual Actions for Risk Assessment
(Section 1R13)05000482/2009004-03 NCV Inadequate Evaluation of Emergency Diesel Generator
for Common Cause Failure in the Supporting Essential
Service Water System (Section 1R15.1)05000482/2009004-04 NCV Failure to Implement Engineered Safety Features
Actuation System Technical Specifications Results in
Missed Mode Change (Section 1R15.2)05000482/2009004-05 NCV Use of Nonsafety-Related Power to Ensure Operability of
Safety-Related Boric Acid System (Section 1R17)05000482/2009004-06 NCV Performing Prohibited Elective Maintenance on Safety
Bus NB02 Channel 4 During Emergency Diesel
Generator Maintenance (Section 1R19)05000482/2009004-07 NCV Failure to Report Conditions that Could have Presented
Fulfillment of a Safety Function (Section4OA3)05000482/2009004-08 NCV Changes to the Approved Fire Protection Program
Without Prior Staff Approval (Section 4OA5.3)
A-1 Attachment
Closed
05000482/2005008-02 AV Failure to Maintain Reactor Coolant System Subcooling
During the Alternative Shutdown (Section 4OA5.4)05000482/2008010-03 URI Changes to the Approved Fire Protection Program May
Not Meet NRC Acceptance Criteria (Section 4OA5.3)
LIST OF DOCUMENTS REVIEWED
Section 1RO1: Adverse Weather Protection
DOCUMENTS
NUMBER TITLE REVISION
STS NB-005 Breaker Alignment Verification Revision 18
Section 1RO4: Equipment Alignment
DOCUMENTS
NUMBER TITLE REVISION
CKL AL-120 Auxiliary Feedwater Normal Lineup 34
M-12AL01 Piping and Instrumentation Diagram - Auxiliary 10
Feedwater System
M-12EF01 Piping and Instrumentation Diagram - Essential 21
Service Water System
M-12EF02 Piping and Instrumentation Diagram - Essential 25
Service Water System
M-12AB01 Piping and Instrumentation Diagram - Main Steam 11
System
M-12AB02 Piping and Instrumentation Diagram - Main Steam 12
System
A-2 Attachment
Section 1RO4: Equipment Alignment
DOCUMENTS
NUMBER TITLE REVISION
USAR 15.6-12/13, Steam Generator Tube Rupture 22
with Postulated Stuck-Open Atmospheric Relief Valve
Control Room Logs dated September 16, 2009 at
1:49 a.m.
M-224A-00037 10 -900 Carbon Steel Flex Wedge Gate Valve with 6:1 G
B.G. Actuator
USAR Figure 9.3-8-03, Piping and Instrumentation 41
Diagram Chemical & Volume Control System
Condition Reports
00019813 00019821 00019825
Work Order
09-3160637-000
Work Request
09-072489
Section 1RO5: Fire Protection
DOCUMENTS
NUMBER TITLE REVISION
AP 10-106 Fire Preplans 8
FPPM-009 Control Bldg El.2000 2
FPPM-014 Diesel Generator Rooms El.2000 1
A-3 Attachment
Section 1R11: Licensed Operator Requalification Program
DOCUMENTS
NUMBER TITLE REVISION /
DATE
LR5004004 Shutdown LOCA 009
Section 1R13: Maintenance Risk Assessment and Emergent Work Controls
DOCUMENTS
NUMBER TITLE REVISION /
DATE
Week of 3/16/09 - Operational Risk Assessment
Condition Reports
00016735 00015318 2009-001338
Section 1R15: Operability Evaluations
DOCUMENTS
NUMBER TITLE REVISION /
DATE
STS SF002 Core Axial Flux difference 9
STS RE-009 Heat Flux Hot Channel Factor Measurement 14
SYS SR-200 Moveable Incore Detector Operation 21
STS RE-012 QPTR Determination 10
STS RE-013C BEACON SinglePoint AFD Calibration 10
WO 09-318203-002 Engineering Disposition: EF138HBC-30 has a thru June 30, 2009
wall leak
WO 09-318203-009 Engineering Disposition: Minimum Wall Issues with July 16, 2009
Line EF138HBC-30
A-4 Attachment
Section 1R15: Operability Evaluations
DOCUMENTS
NUMBER TITLE REVISION /
DATE
SWO 09-318982-001 Engineering Disposition: EF150HBC-18 Essential July 28, 2009
Service Water Pipe Pit Through Wall Leak
WO 09-319429-001 Engineering Disposition: EF049HBC-8 Thru Wall Leak August 20, 2009
Evaluation
SYS KJ-200 Inoperable Emergency Diesel 15 / June 30,
2009
NRC GL 93-05 Line-Item Technical Specifications Improvements to September 27,
Reduce Surveillance Requirements for Testing During 1993
Power Operation
NPF-42 Amendment #101 August 9, 1996
NPF-42 Amendment #123 December 31,
1999
STS IC-232 Channel Operational Test Nuclear Instrumentation 15
System Source Range N-32 Protection Set II
Class IE Environmental Qualification Data Sheet for NI September 1990
31 and 32 Source Range Monitors
Condition Reports
00018217 00018347 00018611 00018945 00019276
00019282 00019307
Work Orders
09-318203-001 (Ultrasonic Thickness Report)
09-318268-000 (Ultrasonic Thickness Report)
09-318269-000 (Ultrasonic Thickness Report)
09-318270-000 (Ultrasonic Thickness Report)
09-318271-000 (Ultrasonic Thickness Report)
09-318272-000 (Ultrasonic Thickness Report)
09-318982-000 (Ultrasonic Thickness Report)
09-318982-003 (Ultrasonic Thickness Report)
09-318982-013 (2 Ultrasonic Thickness Reports)
A-5 Attachment
09-318982-014 (3 Ultrasonic Thickness Reports)
09-319473-000 (Ultrasonic Thickness Report)
09-319473-001 (Ultrasonic Thickness Report)
09-319473-002 (Ultrasonic Thickness Report)
09-319473-003 (Ultrasonic Thickness Report)
09-319473-004 (Ultrasonic Thickness Report)
09-319473-006 (Ultrasonic Thickness Report)
09-319473-007 (Ultrasonic Thickness Report)
09-319473-008 (Ultrasonic Thickness Report)
09-319473-009 (Ultrasonic Thickness Report)
09-319473-010 (Ultrasonic Thickness Report)
Section 1R17: Permanent Plant Modifications (71111.17A)
Calculations
Number Title Revision
M-628-00131-W01 Control Logic Diagram MSIV PPS-700 0
M-630-0095-W01 Control Logic Diagram MFIV PPS-300 0
XX-E-013 Post-Fire Safe Shutdown Analysis 1
XX-E-016 XNB02 Tap Change Analysis 0
XX-E-006 AC System Analysis 5
Wolf Creek Generating Station Rod Withdrawal at Power
AN-06-007 0
(RWAP) Event Safety Analysis
AN-04-015 Radiological Consequences of a Fuel Handling Accident 1
Wolf Creek Generating Station (WCGS) Simplified Head
0720517.01-C-001 0
Assembly (SHA) Drop Analysis
Installation of Vent Lines on Check Valves EJ8958A,
EJ-S-008 0
EJ8958B and EJ8958C
Westinghouse Class I Nuclear Valves 6 and Larger
XX-S-036 0
Swing Check Valves - EM5093
Condition Reports
2006-000363 2006-000577 2006-001070 2006-001447
2006-001858 2006-001923 2006-002412 2006-003067
2006-003135 2006-003235 2006-003241 2007-000070
2007-000235 2007-000416 2007-001115 2007-002153
2007-002251 2007-002329 2007-002401 2007-002459
2007-002727 2007-003578 2007-003767 2007-003782
A-6 Attachment
2007-004696 2008-000028 2008-000083 2008-000662
2008-000826 2008-001445 2008-001727 2008-002157
2008-004744 2008-005500 2008-005550 2008-005808
2009-000409 00014799 00016231 2007-001180
2006-000309 2006-000442 2006-001447 2007-001457
2006-001549 2006-003684
Drawings
Number Title Revision
WIP-E-15000- Electrical Cable, Termination, and Raceway List 5
065-R-1
E-13AB32 Miscellaneous Circuits 7
E-11025 Relay Settings Tabulation and Coordination Curves System 13
NE
0405-0003-01 Intercooler Heat Exchanger Analysis Input Data 2
Miscellaneous
Number Title Date/Revision
USA 50.59 Resource Manual 3
NEI 96-07 Guidelines for 10 CFR 50.59 Implementation 1
J-200B-00001 Nutherm Qualification Report Eaton Cutler-Hammer 0
Contact Blocks With Separation Barriers
PSA PCR 2006- Maintenance of the Wolf Creek PSA Model 0
0002
PSA 05-0002 WCGS PRA Initiating Event Notebook - 2002 Update 0
M-018B-00001 Instruction Manual for Governor Modification W03
Design Change Process Improvements Engineering
N/A 0
Initiative Plan
Design Change Process Improvement Initiative: Monthly
N/A April 10, 2009
Progress Report
A-7 Attachment
Procedures
Number Title Revision
GEN 00-004 Power Operation 54
OFN RP-017 Control Room Evacuation 29
SYS EP-200 Safety Injection Accumulator Operations 30
AP 05-001 Change Package Planning and Implementation 7
AP 05-002 Dispositions and Change Packages 8
Design, Implementation & Configuration Control of
AP 05-005 13
Modifications
AP 05F-001 Design Verification 3
AP 26A-003 10 CFR 50.59 Reviews 10
Section 1R18: Plant Modifications
NUMBER TITLE DATE
09-005-XX-01 Temporary Modification Order February 19, 2009
09-008-SG00 Temporary Modification Order March 5, 2009
09-0019 Essential Required Reading: Responding to an March 12, 2009
Earthquake with Inoperable Seismic Instrumentation
Change Package No. 011613
Condition Reports
2009-001278 2009-001194
Work Requests
09-072504 09-072505 09-072506 09-072507 09-072508
A-8 Attachment
Section 1R19: Postmaintenance Testing
DOCUMENTS
NUMBER TITLE REVISION /
DATE
STS KJ-015B Manual/Auto Fast Start, Sync & Loading of EDG NE02 27A /
August 17,
2009
STS AL-103 Turbine Driven Auxiliary Feedwater Pump Inservice Pump 44 /
Test September
9, 2009
NP-1490 4-900 ANSI Trip Throttle Valve A
103171D Trip Throttle Valve Electrical Schematic Sheet 1 June 5,
1977
103171D Trip Throttle Valve Electrical Schematic Sheet 2 November
17, 1980
Work Orders
09-316773-000 09-316773-001
Section 1R20: Refueling and Other Outage Activities
DOCUMENTS
NUMBER TITLE REVISION
Feedwater Isolation Logic Drawing
Table 7.3.15 USAR - NSSS Interlocks for Engineered Safety Feature 13
Actuation System
SYS SB-122 Enabling/Disabling P-4/LO Tavg Fwis 1
Table 7.5-1 Engineered Safety Features - Displays 21
12.2-7 Westinghouse Technology Systems Manual Reactor 0100
Protection system - Reactor Trip Signals
A-9 Attachment
SYS SB-122 Enable/Disabling P-4/LO Tavg FWIS 1
7.2-31 USAR - Testing of Reactor Trip Breakers 11
Work Orders
09-314863-002 09-319404-000
Performance Improvement Request
2001-0041
Condition Reports
00019318 00019318
Section 1R22: Surveillance Testing
PROCEDURES
NUMBER TITLE REVISION /
DATE
STS CR-004 Shift Log for Additional Monitoring 0
STS EF-100B ESW System Inservice Pump B & ESW B Discharge Check 32 / August
Valve Test 13, 2009
STS IC-208B 4 kV Loss of Voltage and Degraded Voltage TADOT NB02 2A /July 14,
Bus - Separation Group 4 2009
STS RE-006 End of Life Core Moderator Temperature Coefficient 18 / August
Measurement 28, 2009
Work Order
09-315436-000
Condition Reports
00019069 00019000
A-10 Attachment
Section 4OA1: Performance Indicator Verification
PROCEDURES
NUMBER TITLE REVISION
STS IC-203 Channel Operational Test 7300 Process Instrumentation 22B
Protection Set III (Blue)
INC C-001 7300 Signal Comparator Card (NAL 1) 6
MISCELLANEOUS DOCUMENTS
NUMBER TITLE REVISION
NEI 99-02 Mitigating Systems Cornerstone 5
LER 2005-004-00 Failure of Auxiliary Building Ventilation Dampers to
Close on Safety Injection Signal
OPR01 Operability/Reportability Detail Report
LER 2008-007-00 Two Residual Heat Removal Trains Inoperable in
Mode 3 due to Check Valve Leakage
LER 2008-004-00 Loss of Power Event When the Reactor was De-
fueled
LER 2008-008-01/02 Potential for Residual Heat Removal Trains to Be
Inoperable During Mode Change
LER 2008-009-00 Inadequate Compensatory Actions for a Fire Area
LER 2008-001-00 Containment Cooler Inoperability (Callaway Plant
Unit 1)
AIF 16C-001-02 Maintenance Walkdown Form (Technician) 0
10466-M-761-2076- Interconnecting Wiring Diagram Cabinet 03 SNUPPS
W05 Nuclear Power Plant Controls
2000801894 Adverse Condition- Ameren
A-11 Attachment
MISCELLANEOUS DOCUMENTS
NUMBER TITLE REVISION
NCV Failure to Ensure the Suitability of the Design of the
05000483/2008003-01 Containment Air Cooler control Circuitry
Appendix D, 10 CFR 50.72 Including Statement of
Considerations
Event Notification Report of June 23, 2008
Condition Reports
2009-00017786 2009-00017846 2009-00017851 2009-00019914 2009-00019371
2009-001326 2009-00017776 2009-0001261 2009-001326 2009-001004
2008-001307 2009-00018156 200-00018156 2008-000470 2008-001673
Work Orders
09-314726-000 09-317948-000 09-306203-000
Corrective Action Plan
4160 1970 3944 3943
Reportability Evaluation Request
2008-011 2009-012
Section 4OA2: Identification and Resolution of Problems
PROCEDURES
NUMBER TITLE REVISION /
DATE
TMP 09-014 CCW Flow Balance for Troubleshooting Thermal Barrier 0 / July 15,
Closure 2009
SYS EG-201 Transferring Supply of CCW Service Loop and CCW Train 36 / July 15,
Shutdown 2009
A-12 Attachment
Applicability Determination for TMP 09-014 July 14,
2009
50.59 Screen for TMP 09-014 July 14,
2009
Work Order
09-316483-000
Corrective Action
00018793
Section 4OA5: Other Activities
PROCEDURES
NUMBER TITLE DATE
SLNRC 84-109 Letter to NRC 08/23/1984
USAR CR Updated Final Safety Analysis Report Change Request 09/25/2008
2008-009
Evaluation of Proposed Change for USAR CR 2008-009 09/25/2008
Performance Improvement Request
2008-004869
A-13 Attachment