ML093140803

IR 05000482-09-004, 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

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)

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 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

Emergency Preparedness

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.

-8- Enclosure

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.

H.1.a]

- 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 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 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

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

A-13 Attachment