Text

IR 05000482-08-002, on 1/01 - 4/07/08, Wolf Creek Generating Station, Fire Protection, Maintenance Risk Assessments and Emergent Work Control, Access Control to Radiologically Significant Areas, Event Followup and Other Activities
	ML081430159
Person / Time
Site:	Wolf Creek
Issue date:	05/21/2008
From:	Vincent Gaddy; NRC/RGN-IV/DRP/RPB-B
To:	Muench R; Wolf Creek
References
	EA-08-052 IR-08-002
	Download: ML081430159 (65)
	v • d • e

UNITED STATES

NUC LE AR RE G UL AT O RY C O M M I S S I O N

R E GI ON I V

612 EAST LAMAR BLVD , SU I TE 400

AR LI N GTON , TEXAS 76011-4125

May 21, 2008

EA-08-052

Rick A. Muench, President and

Chief Executive Officer

Wolf Creek Nuclear Operating Corporation

P.O. Box 411

Burlington, KS 66839

SUBJECT: WOLF CREEK GENERATING STATION - NRC INTEGRATED INSPECTION

REPORT 05000482/2008002

Dear Muench:

On April 7, 2008, the U.S. Nuclear Regulatory Commission (NRC) completed an integrated

inspection at your Wolf Creek Generating Station. The enclosed report documents the

inspection results, which were discussed on April 11, 2008, with Mr. Stephen Hedges and other

members of your staff.

The inspection examined activities conducted under your license as they relate to safety and

compliance with the Commissions rules and regulations and with the conditions of your license.

The inspectors reviewed selected procedures and records, observed activities, and interviewed

personnel.

Based on the results of this inspection, six NRC identified and two self-revealing findings of very

low safety significance (Green) are documented in this report. All of these findings were

determined to involve violations of NRC requirements. Additionally, three licensee-identified

violations of very low safety significance is listed in this report. However, because of the very

low safety significance and because the findings were entered into your corrective action

program, the NRC is treating these violations as noncited violations consistent with Section VI.A

of the NRC Enforcement Policy.

If you contest these noncited violations, you should provide a response within 30 days of this

inspection report, with the basis for your denial, to the U.S. Nuclear Regulatory Commission,

ATTN.: Document Control Desk, Washington, DC 20555-0001; with copies to the Regional

Administrator Region IV; the Director, Office of Enforcement, U.S. Nuclear Regulatory

Commission, Washington, DC 20555-0001; and the NRC Resident Inspector at the Wolf Creek

Generating Station.

Wolf Creek Nuclear Operating Corp. -2-

In accordance with 10 CFR 2.390 of the NRCs Rules of Practice, a copy of this letter and its

enclosure will be made available electronically for public inspection in the NRC

Public Document Room or from the Publicly Available Records component of NRCs document

system (ADAMS), accessible from the NRC Web site at

http://www.nrc.gov/reading-rm/adams.html (the Public Electronic Reading Room).

Sincerely,

/RA/

Vincent G. Gaddy, Chief

Project Branch B

Division of Reactor Projects

Docket No. 50-482

License No. NPF-42

Enclosure: Inspection Report 05000482/2008002

w/Attachment: Supplemental Information

cc w/enclosure:

Vice President Operations/Plant Manager

Wolf Creek Nuclear Operating Corp.

P.O. Box 411

Burlington, KS 66839

Jay Silberg, Esq.

Pillsbury Winthrop Shaw Pittman LLP

2300 N Street, NW

Washington, DC 20037

Supervisor Licensing

Wolf Creek Nuclear Operating Corp.

P.O. Box 411

Burlington, KS 66839

Chief Engineer

Utilities Division

Kansas Corporation Commission

1500 SW Arrowhead Road

Topeka, KS 66604-4027

Office of the Governor

State of Kansas

Topeka, KS 66612

Wolf Creek Nuclear Operating Corp. -3-

Attorney General

120 S.W. 10th Avenue, 2nd Floor

Topeka, KS 66612-1597

County Clerk

Coffey County Courthouse

110 South 6th Street

Burlington, KS 66839-1798

Chief, Radiation and Asbestos

Control Section

Kansas Department of Health and

Environment

Bureau of Air and Radiation

1000 SW Jackson, Suite 310

Topeka, KS 66612-1366

Ronald L. McCabe, Chief

Technological Hazards Branch

National Preparedness Division

DHS/FEMA

9221 Ward Parkway

Suite 300

Kansas City, MO 64114-3372

Wolf Creek Nuclear Operating Corp. -4-

Electronic distribution by RIV:

Regional Administrator (Elmo.Collins@nrc.gov)

DRP Director (Dwight.Chamberlain@nrc.gov)

DRS Director (Roy.Caniano@nrc.gov)

DRS Deputy Director (Troy.Pruett@nrc.gov)

Senior Resident Inspector (Steve.Cochrum@nrc.gov)

SRI, Callaway (David.Dumbacher@nrc.gov)

Branch Chief, DRP/B (Vincent.Gaddy@nrc.gov)

Senior Project Engineer, DRP/B (Rick Deese@nrc.gov)

Team Leader, DRP/TSS (Chuck.Paulk@nrc.gov)

RITS Coordinator (Marisa.Herrera@nrc.gov)

Only inspection reports to the following:

DRS STA (Dale.Powers@nrc.gov)

J. Adams, OEDO RIV Coordinator (John.Adams@nrc.gov)

P. Lougheed, OEDO RIV Coordinator (Patricia.Lougheed@nrc.gov)

R. Kahler, NSIR/DPR/EDP (Robert.Kahler@nrc.gov)

ROPreports Resource

OEMail Resource

WC Site Secretary (Shirley.Allen@nrc.gov)

SUNSI Review Completed: __VGG__ ADAMS: ; Yes No Initials: __VGG__

Publicly Available Non-Publicly Available Sensitive ; Non-Sensitive

R:\_REACTORS\_WC\2007\WC2008-002RP-SDC.wpd ADAMS ML081430159

SRI:DRP/B RI:DRP/B C:DRS/EB1 C:DRS/EB2

SDCochrum CMLong RBywater LJSmith

/RA/ /RA/ /RA/ /RA/ GAPick for

5/21/2008 5/21/2008 4/28/2008 4/30/2008

C:DRS/OB C:DRS/PSB ACES/SES C:DRP/B

RELantz MPShannon MMVasquez VGGaddy

/RA/ /RA/ /RA/ /RA/

4/29/2008 4/28/2008 5/13/2008 5/21/2008

OFFICIAL RECORD COPY T=Telephone E=E-mail F=Fax

U. S. NUCLEAR REGULATORY COMMISSION

REGION IV

Docket: 50-482

License: NPF-42

Report: 5000482/2008002

Licensee: Wolf Creek Operating Corporation

Facility: Wolf Creek Generating Station

Location: 1550 Oxen Lane NE

Burlington, Kansas

Dates: January 1 to April 7, 2008

Inspectors: S. D. Cochrum, Senior Resident Inspector

C. M. Long, Resident Inspector

G. A. Pick, Senior Reactor Inspector

D. L. Stearns, Health Physics Inspector

Approved by: V. G. Gaddy, Chief, Project Branch B

-1- Enclosure

TABLE OF CONTENTS

SUMMARY OF FINDINGS ......................................................................................................... 3

REPORT DETAILS..................................................................................................................... 9

REACTOR SAFETY ................................................................................................................... 9

1R01 Adverse Wather Protection (71111.01)..........................................................................9

1R04 Equipment Alignment (71111.04) ............................................................................. 10

1R05 Fire Protection (71111.05) ......................................................................................... 11

1R11 Licensed Operator Requalification Program (71111.11) ............................................ 13

1R12 Maintenance Effectiveness (71111.12)...................................................................... 14

1R13 Maintenance Risk Assessments and Emergent Work Control (71111.13) ................. 14

1R15 Operability Evaluations (71111.15)............................................................................ 17

1R19 Postmaintenance Testing (71111.19) ........................................................................ 19

1R20 Outage Activities (71111.20) ..................................................................................... 20

1R22 Surveillance Testing (71111.22) ............................................................................... 22

1EP6 Drill Evaluation (71114.06) ........................................................................................ 24

2OS1 Access Control to Radiologically Significant Areas (71121.01) ..24

2OS2 ALARA Planning and Controls (71121.02)................................................................. 28

OTHER ACTIVITIES ................................................................................................................ 29

4OA1 Performance Indicator Verification (71151)................................................................ 29

4OA2 Identification and Resolution of Problems (71152)..................................................... 32

4OA3 Event Followup .......................................................................................................... 34

4OA5 Other Activities .......................................................................................................... 44

4OA6 Meetings, Including Exit............................................................................................. 47

4OA7 Licensee-Identified Violations .................................................................................... 47

SUPPLEMENTAL INFORMATION ..........................................................................................A-1

KEY POINTS OF CONTACT ...................................................................................................A-1

ITEMS OPENED, CLOSED, AND DISCUSSED ......................................................................A-1

LIST OF DOCUMENTS REVIEWED .......................................................................................A-2

LIST OF ACRONYMS ...........................................................................................................A-16

-2- Enclosure

SUMMARY OF FINDINGS

IR 05000482/2008002; 1/01 - 4/07/08; Wolf Creek Generating Station; Fire Protection,

Maintenance Risk Assessments and Emergent Work Control, Access Control to Radiologically

Significant Areas, Event Followup and Other Activities.

This report covered a 3-month period of inspection by resident inspectors and regional

specialists. The inspection identified eight Green findings, all of which are noncited violations.

The significance of most findings is indicated by their color (Green, White, Yellow, or Red) using

Inspection Manual Chapter 0609, "Significance Determination Process." Findings for which the

significance determination process does not apply may be Green or be assigned a severity level

after NRC managements review. The NRCs program for overseeing the safe operation of

commercial nuclear power reactors is described in NUREG-1649, Reactor Oversight Process,

Revision 4, dated December 2006.

A. NRC-Identified and Self-Revealing Findings

Cornerstone: Mitigating Systems

Green. The inspectors identified a noncited violation of Technical

Specification 5.4.1.d for failure to implement fire protection impairment control

permit requirements and compensatory measures when operators received a

trouble alarm on a fire detector in the auxiliary building. On January 26, 2008,

operators discovered that Detector KC-104-XCH-ID-006 did not have a fire

protection impairment control permit. This detector was adjacent to Detector KC-

104-XSH-ID-007 which was already inoperable in Impairment 2008-020. The

licensees administrative procedure required fire detection in the area to be

declared inoperable if two adjacent detectors are inoperable. This condition

existed for approximately 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months and would have required a compensatory

continuous fire watch for the period that both detectors were inoperable. The

residents identified that the control room turnover checklist contains a section for

listing the KC008 alarms; however, the two turnover checklists for the two shifts

following the initial alarm did not identify Detector KC-104-XCH-ID-006 as a

Detector KC-008 alarm.

The failure to implement fire protection impairment control permit requirements

and establish compensatory measures for the auxiliary building 2026-foot level

was considered a performance deficiency. The finding was more than minor

because it was associated with the Mitigating Systems Cornerstone attribute of

protection against external factors and affected the cornerstone objective of

ensuring the availability of systems that respond to initiating events to prevent

undesirable consequences. Specifically, this issue relates to the protection

against fire example of protection against external factors attribute because the

detectors were inoperable without ensuring compensatory measures were in

place. The finding was of very low safety significance because it involved

compensatory measures for the fixed fire protection system and was assigned a

low degradation rating since less than 10 percent of the fire detectors in the area

were disabled. The finding has crosscutting aspects in the area of human

performance associated with work practices because the licensee failed to apply

appropriate human error techniques such as self and peer-checking techniques

to avoid committing errors H.4(a) (Section 1R05).

-3- Enclosure

Green. A noncited violation of Technical Specification 3.8.1.B.4 was identified

when the licensee performed elective maintenance in the switchyard and

removed equipment from service that was prohibited by Technical Specifications

while in an extended diesel generator outage.

The inspectors determined that the failure to implement requirements of

Technical Specification 3.8.1.B.4 was a performance deficiency. The finding was

more than minor because it is associated with the equipment performance

attribute for the mitigating systems cornerstone; and, it affected the cornerstone

objective to ensure the availability, reliability, and capability of systems that

respond to initiating events to prevent undesirable consequences (i.e. core

damage). The finding was determined to be of very low safety significance

because the issue resulted in the Train B offsite power being inoperable, but

capable of supplying the safety bus for greater than 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . Additionally, the

cause of the finding has a human performance crosscutting aspect in the area

associated with work control. Specifically, Wolf Creek did not ensure

STS-IC-805B was appropriately coordinated within organizations to assure plant

and human performance during the extended emergency diesel generator

allowed outage time H.3(b) (Section 1R13).

Green. The NRC identified a noncited violation of Technical Specification 5.4.1

for failure to follow the operability process on discovery of the centrifugal

charging Pump A room cooler leak. On February 13, 2008, at 2:20 p.m., the

control room was notified of a leak from the room cooler for the centrifugal

charging pump. At that time, it could not be established if the leak would cause a

loss of structural integrity of essential service water. Wolf Creek made no log

entries at 2:20 p.m. stating its basis for immediate operability. At 3:50 p.m., Wolf

Creek control room logs documented that centrifugal charging Pump A had a

room cooler leak and structural integrity cannot be verified. Subsequent entry

into Technical Specification 3.7.8 for the essential service water Pump A caused

emergency diesel Generator A to be inoperable. Technical Specification 3.8.1,

Condition I, states that with three alternating current sources inoperable (both

emergency diesel generators and an offsite source), Technical

Specification 3.0.3 shall be entered. Wolf Creek exited Technical

Specification 3.0.3 at 4:13 p.m. when the inlet and outlet valves to centrifugal

charging Pump As room cooler were closed. The inspectors could not locate

any justification produced by Wolf Creek for the room coolers operability after

2:20 p.m.

The inspectors determined that the failure to follow the operability process is a

performance deficiency. The inspectors determined that this finding was more

than minor because if left uncorrected, it could become a more serious problem if

the operability process is not correctly applied. The finding screened to Phase 2

because the finding represents an actual loss of safety function of a single train

of high head injection. A bounding risk of Green results from the Phase 2

presolved worksheets using an exposure time of less than 3 days for the

Centrifugal Charging Pump (CCP) A [Fails to Run]. The inspectors also

determined that the finding had a human performance crosscutting aspect in the

area associated with decision making because the licensee failed to use

conservative assumptions in its operability decision and apply a requirement to

-4- Enclosure

demonstrate that the room cooler is operable in order to proceed rather than a

requirement to demonstrate that it is inoperable H.1(b) (Section 4OA3.2(ii)).

Green. The inspectors identified a noncited violation of 10 CFR Part 50

Appendix B Criterion XVI, Corrective Action, because Wolf Creek failed to take

timely corrective actions to prevent failure of the centrifugal charging pump A

room cooler which resulted in a Notice of Enforcement Discretion (EA-08-052).

The inspectors found that room Cooler SGL12A experienced leaks in

October 1999, May 2003, October 2003, August 2004, October 2006, and again

in February 2008. On March 14, 2007, Wolf Creek chose to delay SGL12As

replacement until Refueling Outage 16 due to the required length of time to

replace the cooler. On February 13, 2008, a circumferential flaw on an H-bend

was discovered in SGL12A preventing it from performing its safety function.

Inspectors reviewed corrective action Procedure AP 28A-100, Condition

Reports, Revision 3 and found that a loss of a train to perform its safety function

was considered a significant deficiency requiring corrective action to prevent

recurrence. The inspectors reviewed this issue under Performance Improvement

Requests 2005-2507 and 2004-0688, and Condition Report 2008-0467 and

found that Wolf Creek designated prior failures nonsignificant.

The failure to take timely corrective actions within 9 years was a performance

deficiency. The inspectors determined that this finding was more than minor

because it is associated with the equipment performance attribute for the

mitigating systems cornerstone; and, it affected the cornerstone objective to

ensure the availability, reliability, and capability of systems that respond to

initiating events to prevent undesirable consequences (i.e. core damage). The

finding screened to Phase 2 because the finding represents an actual loss of

safety function of a single train of high head injection for greater than its

Technical Specification 3.8.1.B.2 allowed outage time of 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . Using an

exposure time of less than 3 days for the scenario Centrifugal Charging Pump

PBG05A [Fails to Run], a bounding risk of Green results from the Phase 2

presolved worksheets. Additionally, the cause of the finding has a human

performance crosscutting aspect in the area associated with resources.

Specifically, Wolf Creek did not ensure adequate resources to maintain long-term

plant safety by minimizing the room coolers long-standing issues and preventive

maintenance deferrals H.2(a) (Section 4OA3.2(iii)).

Green. The inspectors identified a noncited violation of Technical Specification 5.4.1.d because Procedure OFN RP 017, "Control Room Evacuation," Revision

21, failed to account for the needed actions to reestablish reactor coolant pump

seal cooling. Failure to reestablish seal cooling in a timely manner could have

resulted in a small break loss of coolant accident.

This performance deficiency resulted from an inadequate postfire safe shutdown

procedure. The inspectors determined the finding is greater than minor in that it

affected the ability to achieve and maintain hot shutdown following a control room

fire. This finding is associated with the mitigating systems cornerstone attribute

of protection against external factors (e.g. fire). This finding affected the

mitigating systems cornerstone objective to ensure the availability, reliability, and

capability of systems that respond to external events (such as fire) to prevent

undesirable consequences. In addition to the control room fire requiring

-5- Enclosure

operators to evacuate the control room, the fire would have had to affect

components located in two physically separated panels. The licensee has

Institute of Electrical and Electronics Engineers Standard 383 qualified cables

and conductors throughout the plant. The Phase 3 risk evaluation performed by

the NRC senior reactor analyst determined this deficiency had very low risk

significance (Section 4OA3.5).

Green. The inspectors identified a noncited violation of License Condition 2.c(5)

because the licensee failed to evaluate the impact of a motor-operated valve

failure mechanism on their ability to implement postfire safe shutdown following a

control room evacuation. The licensee determined that the failure mechanism

affected 38 motor-operated valves and upon valve failure could affect their ability

to implement their postfire safe shutdown procedure. A short circuit that

bypassed the torque and/or limit switches could damage the valves and prevent

repositioning of the valve in the postfire safe shutdown position.

The inspectors determined this was a performance deficiency because the

licensee failed to ensure that components necessary to safely shutdown the

reactor would remain operable following a fire. This deficiency was more than

minor, in that, it had the potential to impact the mitigating systems cornerstone

objective to ensure the availability, reliability, and capability of systems that

respond to external events (such as fire) to prevent undesirable consequences.

In addition to the control room fire requiring operators to evacuate the control

room, the fire would have had to affect components located in five different

control panels. The Phase 3 risk evaluation performed by the NRC senior reactor

analyst determined this deficiency had very low risk significance

(Section 4OA5.2).

Cornerstone: Occupational Radiation Safety

Green. The inspectors reviewed a self-revealing noncited violation of Technical

Specification 5.7.2.a for failure to evaluate changing radiological conditions and

control an area as a locked high radiation area. Specifically, on October 17,

2007, dose rates in Room 7604 increased to levels requiring posting as a

Locked High Radiation Area, as a result of a vent and drain evolution. Dose

rates reached a level of 1500 mRem/hour prior to the area being properly posted

and controlled. This issue was entered into the licensees corrective action

program as Condition Report 2007-003934. Immediate corrective actions

included posting and controlling the area as a locked high radiation area. Other

corrective actions included changing the vent and drain process to change the

vent path.

This finding is greater than minor because it is associated with the occupational

radiation safety program and process attribute and affected the cornerstone

objective, in that, the failure to properly post and control access to a locked high

radiation area has the potential to increase personnel dose. This occurrence

involves the potential for unplanned, unintended dose. Utilizing Inspection

Manual Chapter 0609, Appendix C, Occupational Radiation Safety Significance

Determination Process, the inspectors determined that the finding was of very

low safety significance because it did not involve; (1) as low as is reasonably

achievable planning and controls, (2) an overexposure, (3) a substantial potential

-6- Enclosure

for overexposure, or (4) an impaired ability to assess dose. This finding has a

crosscutting aspect in the area of human performance associated with the work

control component because licensee failed to appropriately plan work activities

by incorporating job site conditions that may impact radiological safety H.3(a)

(Section 2OS1(i)).

Green. The inspectors reviewed a self-revealing noncited violation of Technical

Specification 5.4.1 for failure to follow a licensee procedure. Specifically, on

March 29, 2008, one of two radiographers conducting radiography operations in

the quality control vault received a dose rate alarm on their electronic dosimeter.

The two radiographers evaluated the dose received and decided to continue with

radiography without notifying health physics personnel to evaluate the conditions.

This issue was entered into the licensees corrective action program as Condition

Report 2008-001181. Immediate corrective actions included restriction of the

radiographers to log onto the radiation work permit and discussions with the

radiographers and the contractors radiation safety officer. Long-term corrective

action is still being evaluated.

This finding is greater than minor because it is associated with the occupational

radiation safety program and process attribute and affected the cornerstone

objective, in that, the failure to stop work and notify health physics personnel for

assistance had the potential to increase personnel dose. This occurrence

involves the potential for unplanned, unintended dose. Utilizing Inspection

Manual Chapter 0609, Appendix C, Occupational Radiation Safety Significance

Determination Process, the inspector determined that the finding was of very low

safety significance because it did not involve: (1) as low as is reasonably

achievable planning and controls, (2) an overexposure, (3) a substantial potential

for overexposure, or (4) an impaired ability to assess dose. This finding has a

crosscutting aspect in the area of human performance associated with the

decision making component because the radiographer and assistant failed to

contact health physics personnel to discuss the circumstances surrounding the

unexpected dose rate alarm H.1(a) (Section 2OS1(ii)).

B. Licensee-Identified Violations

Violations of very low safety significance which were identified by the licensee have

been reviewed by the inspectors. Corrective actions taken or planned by the licensee

have been entered into the licensee's corrective action program. These violations and

their corrective actions are listed in Section 4OA7 of this report.

-7- Enclosure

REPORT DETAILS

Summary of Plant Status

The plant started the inspection period at 100 percent rated thermal power. On January 11,

2008, the licensee performed a reactor shutdown due to voiding in emergency core cooling

system (ECCS) piping. After determining the cause and restoring systems to operable, the

plant was returned to full power on January 16, 2008. On March 17, 2008, a 13.8 kV

transformer failure resulted in a plant trip. The plant remained shut down the rest of the report

period and entered Refueling Outage 16 on March 22, 2008.

1. REACTOR SAFETY

Cornerstones: Initiating Events, Mitigating Systems, Barrier Integrity, Emergency

Preparedness

1R01 Adverse Weather Protection (71111.01)

.1 Readiness for Seasonal Susceptibilities

a. Inspection Scope

The inspectors completed a review of the licensee's readiness of seasonal

susceptibilities involving extreme high temperatures. The inspectors: (1) reviewed plant

procedures, the Updated Safety Analysis Report (USAR), and Technical Specifications

to ensure that operator actions defined in adverse weather procedures maintained the

readiness of essential systems; (2) walked down portions of the systems listed below to

ensure that adverse weather protection features were sufficient to support operability

including the ability to perform safe shutdown functions; (3) evaluated operator staffing

levels to ensure the licensee would maintain the readiness of essential systems required

by plant procedures; and (4) reviewed the corrective action program to determine if the

licensee identified and corrected problems related to adverse weather conditions.

January 23, 2008, cold weather impact on essential service water (ESW)

Documents reviewed are listed in the attachment.

The inspectors completed one sample.

b. Findings

No findings of significance were identified.

.2 Readiness for Impending Adverse Weather Conditions

a. Inspection Scope

The inspectors completed a review of the licensee's readiness for impending adverse

weather involving severe thunderstorms and heavy rains. The inspectors: (1) reviewed

plant procedures, the Updated Safety Analysis Report (USAR), and Technical

Specifications to ensure that operator actions defined in adverse weather procedures

-8- Enclosure

maintained the readiness of essential systems; (2) walked down portions of the systems

listed below to ensure that adverse weather protection features were sufficient to support

operability, including the ability to perform safe shutdown functions; (3) reviewed

maintenance records to determine that applicable surveillance requirements were

current before the anticipated weather developed; and (4) reviewed plant modifications,

procedure revisions, and operator work arounds to determine if recent facility changes

challenged plant operation.

January 7, 2008, severe thunderstorms caused the loss of two alert notification

system sirens

Documents reviewed are listed in the attachment.

The inspectors completed one sample.

b. Findings

No findings of significance were identified.

1R04 Equipment Alignment (71111.04)

.1 Quarterly Partial System Walkdowns

a. Inspection Scope

The inspectors performed partial system walkdowns of the following risk-significant

systems:

February 11, 2008, Sharpe station alignment to Wolf Creek switchyard

March 6, 2008, emergency diesel Generator (EDG) B while ESW A is inoperable

March 12, 2008, motor-driven auxiliary feedwater Pump A while turbine-driven

auxiliary feedwater (TDAFW) is inoperable

The inspectors selected these systems based on their risk significance relative to the

reactor safety cornerstones at the time they were inspected. The inspectors attempted

to identify any discrepancies that could impact the function of the system, and, therefore,

potentially increase risk. The inspectors reviewed applicable operating procedures,

system diagrams, USAR, Technical Specification requirements, administrative Technical

Specifications, outstanding work orders (WOs), condition reports, and the impact of

ongoing work activities on redundant trains of equipment in order to identify conditions

that could have rendered the systems incapable of performing their intended functions.

The inspectors also walked down accessible portions of the systems to verify system

components and support equipment were aligned correctly and operable. The

inspectors examined the material condition of the components and observed operating

parameters of equipment to verify that there were no obvious deficiencies. The

inspectors also verified that the licensee had properly identified and resolved equipment

alignment problems that could cause initiating events or impact the capability of

mitigating systems or barriers and entered them into the corrective action program with

the appropriate significance characterization.

-9- Enclosure

Documents reviewed are listed in the attachment.

The inspectors completed three samples.

b. Findings

No findings of significance were identified.

1R05 Fire Protection (71111.05)

.1 Routine Resident Inspector Tours (71111.05Q)

a. Inspection Scope

The inspectors conducted fire protection walkdowns which were focused on availability,

accessibility, and the condition of firefighting equipment in the following risk-significant

plant areas:

January 17, 2008, auxiliary building 1988 elevation

January 31, 2008, auxiliary building 2026 elevation

February 7, 2008, control building 2000 elevation

March 10, 2007, turbine building 2037 elevation

The inspectors reviewed areas to assess if the licensee had implemented a fire

protection program that adequately controlled combustibles and ignition sources within

the plant, effectively maintained fire detection and suppression capability, maintained

passive fire protection features in good material condition, and had implemented

adequate compensatory measures for out of service, degraded or inoperable fire

protection equipment, systems, or features in accordance with the licensees fire plan.

The inspectors selected fire areas based on their overall contribution to internal fire risk

as documented in the plants individual plant examination of external events (IPEEE)

with later additional insights, their potential to impact equipment which could initiate or

mitigate a plant transient, or their impact on the plants ability to respond to a security

event. Using the documents listed in the attachment, the inspectors verified that fire

hoses and extinguishers were in their designated locations and available for immediate

use; that fire detectors and sprinklers were unobstructed, that transient material loading

was within the analyzed limits; and fire doors, dampers, and penetration seals appeared

to be in satisfactory condition. The inspectors also verified that minor issues identified

during the inspection were entered into the licensees corrective action program.

Documents reviewed by the inspectors are listed in the attachment.

The inspectors completed four samples

b. Findings

Introduction. The inspectors identified a Green noncited violation (NCV) of Technical

Specification 5.4.1.d for failure to implement fire protection impairment control permit

- 10 - Enclosure

requirements and compensatory measures following a trouble alarm for a fire detector in

the auxiliary building.

Description. On January 25, 2008, a fire protection trouble alarm came in on the fire

alarm control Panel KC-008. The control room supervisor acknowledged the alarm

associated with Detector KC-104-XCH-ID-006. Alarm Procedure ALR KC-008, Fire

Protection Panel KC-008 Alarm Response, Revision 15, Step 4.3.1 requires, in part,

that a fire protection impairment permit be initiated for a smoke detector trouble alarm.

The control room supervisor reviewed the impairment log and found

Impairment 2008-020 for a detector in the same location as the alarm. Based on this

information, the control room supervisor did not initiate an impairment request. Alarming

Detector KC-104-XCH-ID-006 was adjacent to Detector KC-104-XSH-ID-007 that was

listed in Impairment 2008-020. However, the control room supervisor did not verify the

alarming detector point was the same detector point listed in the impairment.

On January 26, 2008, while performing Procedure STN KC-008, Fire Alarm Control

Panel KC-008 Daily Check, Revision 7, which required operators to check KC-008

alarms and trouble points, it was discovered that Detector KC-104-XCH-ID-006 did not

have a fire protection impairment control permit. This detector was adjacent to

Detector KC-04-XSH-ID-07 which was already inoperable in Impairment 2008-020.

Administrative Procedure AP 10-103, Fire Protection Impairment Control, Revision 22,

required fire detection in the area to be declared inoperable if two adjacent detectors are

inoperable. This condition existed for approximately 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months and required a

compensatory continuous fire watch for the period that both detectors were inoperable.

Upon discovery, a continuous fire watch was stationed and emergent work was declared

to clean the dirty detectors.

Procedure AP 21-001, Conduct of Operations, Revision 39, requires, in part, that watch

stations are responsible for reviewing and understanding pertinent information such as

control room turnover checklists for shift relief and turnover. The procedure also states

that shift turnover discussions should include reasons for alarms and status panel lights.

The residents noted that the control room turnover checklist contains a section for listing

the KC008 alarms; however, the two turnover checklists for the two shifts following the

initial alarm did not identify Detector KC-104-XCH-ID-006 as a KC-008 alarm. The

control room turnover checklist also has specific requirements to review fire protection

permits and verify that KC-008 alarms are not disabled or disarmed without

documentation. However, neither review discovered that the alarm did not have an

impairment issued. Based on this information, the residents concluded that the licensee

had two previous opportunities to identify the condition during control room turnovers.

After reviewing the licensees evaluation of the condition, the residents noted that these

aspects were not identified in the evaluations conclusions or corrective actions which

focused on only the initial error performed by the operator. During interviews with

control room operators, the inspectors noted that operators are trained to ask for and get

peer checks for verification of alarms and disabled points but failed to utilize any human

error prevention tools in this instance.

Analysis. The failure to implement fire protection impairment control permit

requirements and establish compensatory measures for the auxiliary building 2026' level

was considered a performance deficiency. Traditional enforcement does not apply since

there were no actual safety consequences or potential for impacting the NRCs

- 11 - Enclosure

regulatory function, and the finding was not the result of any willful violation of NRC

requirements or Wolf Creek procedures. The inspectors determined that the finding was

more than minor because it was associated with the mitigating systems cornerstone

attribute of protection against external factors and affected the cornerstone objective of

ensuring the availability of systems that respond to initiating events to prevent

undesirable consequences. Specifically, this issue relates to the protection against fire

example of protection against external factors attribute because the detectors were

inoperable without ensuring compensatory measures where in place. The inspectors

evaluated the significance of this finding using Phase 1 of Inspection Manual

Chapter (IMC) 0609, Appendix F, "Fire Protection Significance Determination Process,

the inspectors determined that the finding was of very low safety significance because it

involved compensatory measures for the fixed fire protection system and was assigned

a low degradation rating since less than 10 percent of the fire detectors in the area were

disabled. The inspectors also determined that the finding has crosscutting aspects in

the area of human performance associated with work practices because the licensee

failed to apply appropriate human error techniques such as self- and peer-checking

techniques to avoid committing errors H.4(a).

Enforcement. Technical Specification 5.4.1.d requires that written procedures be

established, implemented, and maintained covering activities related to fire protection

program implementation. Administrative Procedure AP 10-103, Fire Protection

Impairment Control, Revision 21, requires, in part, fire protection impairment control

permit shall be prepared in order to determine the appropriate compensatory measures

and track the impairment. Contrary to the above, on January 25, 2008, two fire

detectors were inoperable in the auxiliary building 2026' level without implementing a fire

protection impairment control permit and establishing compensatory measures. This

issue and the corrective actions are being tracked by the licensee in Condition

Report (CR) 2008-001657. Because the finding is of very low safety significance and

has been entered into the corrective action program, this violation is being treated as an

NCV 05000482/2008002-01, Failure to Implement Fire Protection Impairment Control

Permit Requirements and Compensatory Measures.

1R11 Licensed Operator Requalification Program (71111.11)

Resident Inspector Quarterly Review (71111.11Q)

a. Inspection Scope

The inspectors observed testing and training of senior reactor operators and reactor

operators to identify deficiencies and discrepancies in the training, to assess operator

performance, and to assess the evaluator's critique. The training scenario involved:

February 1, 2008, loss of residual heat removal (RHR) during shutdown

conditions

Documents reviewed by the inspectors are listed in the attachment.

The inspectors completed one sample.

- 12 - Enclosure

b. Findings

No findings of significance were identified.

1R12 Maintenance Effectiveness (71111.12)

.1 Routine Quarterly Evaluations (71111.12Q)

a. Inspection Scope

The inspectors reviewed the maintenance activities listed below to: (1) verify the

appropriate handling of structure, system, and component (SSC) performance or

condition problems; (2) verify the appropriate handling of degraded SSC functional

performance; (3) evaluate the role of work practices and common cause problems; and

(4) evaluate the handling of SSC issues reviewed under the requirements of the

maintenance rule, 10 CFR Part 50, Appendix B, and Technical Specifications.

November 27, 2007, service water Pump A trip due to SL41 bus transients

Documents reviewed by the inspectors are listed in the attachment.

The inspectors completed one sample.

b. Findings

No findings of significance were identified.

1R13 Maintenance Risk Assessments and Emergent Work Control (71111.13)

a. Inspection Scope

Risk Assessment and Management of Risk

The inspectors reviewed the licensee's evaluation and management of plant risk for the

maintenance and emergent work activities affecting risk-significant and safety-related

equipment listed below to verify that the appropriate risk assessments were performed

prior to removing equipment for work:

March 17-22, 2008, weekly T-0 risk assessment profile

February 12 and 13, 2008, STS IC-805B degraded grid voltage relay testing

These activities were selected based on their potential risk significance relative to the

reactor safety cornerstones. As applicable for each activity, the inspectors verified that

risk assessments were performed as required by 10 CFR 50.65(a)(4) and were accurate

and complete. When emergent work was performed, the inspectors verified that the

plant risk was promptly reassessed and managed. The inspectors reviewed the scope

of maintenance work, discussed the results of the assessment with the licensee's

probabilistic risk analyst or shift technical advisor, and verified plant conditions were

consistent with the risk assessment. The inspectors also reviewed Technical

Specification requirements and walked down portions of redundant safety systems,

- 13 - Enclosure

when applicable, to verify risk analysis assumptions were valid and applicable

requirements were met.

Emergent Work Control

For the emergent work activities listed below, the inspectors: (1) verified that the

licensee performed actions to minimize the probability of initiating events and maintained

the functional capability of mitigating systems and barrier integrity systems; (2) verified

that emergent work-related activities such as troubleshooting, work planning/scheduling,

establishing plant conditions, aligning equipment, tagging, temporary modifications, and

equipment restoration did not place the plant in an unacceptable configuration; and

(3) reviewed the corrective action program to determine if the licensee identified and

corrected risk assessment and emergent work control problems.

January 11, 2008, shutdown due to ECCS voiding

March 11, 2008, scaffolding installation resulting in reactive load swings

Documents reviewed by the inspectors are listed in the attachment.

The inspectors completed four samples.

b. Findings

Introduction. The inspectors identified a Green NCV of Technical Specification 3.8.1.B.4

in which the licensee removed equipment from service that was prohibited by Technical

Specifications.

Description. On February 11, 2008, the licensee entered TS 3.8.1.B.4.2.2. This

specification allowed an emergency diesel generator to be inoperable for up to 7 days

during an once per cycle extended outage. On February 12 and 13, inspectors

observed surveillance Procedure STS IC-805B, Channel Calibration of NB02 Grid

Degraded Voltage, Time Delay Trip, Revision 11, for testing of the Train B of degraded

voltage relays for offsite power. Offsite power Train B was declared inoperable at

10:09 a.m. on February 12. On February 12, instrumentation and control (I&C)

technicians partially completed STS 805B, but required clarification of their procedure

and secured the test and restored the equipment to operable status. On February 13,

offsite power Train B was again declared inoperable at 7:49 a.m. Inspectors reviewed

Technical Specification Bases 3.8.1.B.4 which prohibits elective maintenance within the

switchyard that would challenge offsite power while in the 7-day emergency diesel

generator extended outage. Inspectors also reviewed the NRC Safety Evaluation

Report (SER) for the 7 day EDG allowed outage time (Technical

Specification 3.8.1.B.4.2.2) and found that Section 4.6.c, states: The offsite power

supply and switchyard conditions are conducive to an extend[ed] DG [completion time],

which includes ensuring that switchyard access is restricted and no elective

maintenance within the switchyard is performed that would challenge the offsite power

availability. Additionally, Condition D of the Technical Specification Bases states that

no equipment or systems assumed to be available for the extended EDG completion

time are removed from service, which includes auxiliary feedwater, component cooling

water, essential service water and their support systems. However, Wolf Creek

removed one train of offsite power degraded voltage relays which affects offsite power to

- 14 - Enclosure

Bus NB02 (Train B) which is a support system for the above equipment. The inspectors

found that Procedure STS IC 805B permits the testing of degraded voltage relays only

while the diesel is out of service. The inspectors determined that this practice is

acceptable when performing offsite power maintenance under Technical Specification 3.8.1.B.4.1, but not Technical Specification 3.8.1.B.4.2.2 due to the

increase in risk for the longer allowed outage period. Procedure STS IC-805B was not

revised subsequent to issuance of License Amendment 163 and permitted the work to

occur. Additionally, Procedure AP 22C-003, Operational Risk Assessment Program,

Revision 13, prohibits elective maintenance within the switchyard that would challenge

offsite power during Technical Specification 3.8.1.B.4.2.2. Wolf Creek appropriately

restricted access to the portion of the switchyard outside the protected area but did not

appropriately restrict work for offsite power inside the protected area. The inspectors

determined that challenges to offsite power can originate with elective maintenance

inside the protected area. Inspectors found that Wolf Creek assessed risk under

10 CFR 50.65 a(4) for this evolution which resulted in elevating risk to yellow during

testing.

Analysis. The inspectors determined that the failure to follow the NRC SER and

Technical Specification Bases for Technical Specification 3.8.1.B.4 was a performance

deficiency. Traditional enforcement does not apply since there were no actual safety

consequences or potential for impacting the NRC's regulatory function, and the finding

was not the result of any willful violation of NRC requirements or Wolf Creek procedures.

The inspectors determined that this finding was more than minor because it is

associated with the equipment performance attribute for the mitigating systems

cornerstone; and, it affected the cornerstone objective to ensure the availability,

reliability, and capability of systems that respond to initiating events to prevent

undesirable consequences (i.e. core damage). Specifically, this issue relates to the

availability and reliability examples of the equipment performance attribute because an

offsite power source was at greater risk of being lost.

The inspectors evaluated the significance of this finding using Phase 1 of IMC 0609,

Appendix A, Significance Determination of Reactor Inspection Findings for At-Power

Situations, and determined that the finding was of very low safety significance because

the issue resulted in the Train B offsite power being inoperable, but capable of supplying

the safety bus for greater than 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . As such, under Phase 1 screening, the

deficiency is not related to a qualification or design deficiency, it did not represent a loss

of safety function for a train or system as defined in the plant specific risk-informed

inspection notebook, and was not related to external events such as fires and floods.

Additionally, the cause of the finding has human performance crosscutting aspects in the

area associated with work control. Specifically, Wolf Creek did not ensure STS IC-805B

was appropriately coordinated within organizations to assure plant and human

performance during the extended EDG allowed outage time H.3(b).

Enforcement. Technical Specification 3.8.1, Condition B.4.2.2, permits one diesel

generator to be inoperable for 7 days provided the limitations articulated in the NRC

SER for License Amendment 163. The NRC SER requires that the offsite power supply

and switchyard conditions be conducive to an extend[ed] diesel generator [completion

time], which includes ensuring that switchyard access is restricted and no elective

maintenance within the switchyard is performed that would challenge the offsite power

availability. The NRC SER also requires support equipment to systems such as auxiliary

feedwater, component cooling water, and ESW to be available. Contrary to the above,

- 15 - Enclosure

on February 12 and 13, 2008, Wolf Creek performed elective maintenance on the

Train B offsite power degraded voltage relays while the Train B emergency diesel

generator was in an extended outage. Because the finding is of very low safety

significance and has been entered into the corrective action program as CR 2008-

001675, this violation is being treated as an NCV, consistent with Section VI.A of the

NRC Enforcement Policy: NCV 05000482/2008002-02, Performing Prohibited Elective

Maintenance on Offsite Power During EDG Maintenance.

1R15 Operability Evaluations (71111.15)

a. Inspection Scope

The inspectors: (1) reviewed plants status documents such as operator shift logs,

emergent work documentation, deferred modifications, and standing orders to determine

if an operability evaluation was warranted for degraded components; (2) referred to the

USAR and design basis documents to review the technical adequacy of licensee

operability evaluations; (3) evaluated compensatory measures associated with

operability evaluations; (4) determined degraded component impact on any Technical

Specifications; (5) used the significance determination process to evaluate the risk

significance of degraded or inoperable equipment; and (6) verified that the licensee has

identified and implemented appropriate corrective actions associated with degraded

components.

January 22, 2008, containment sump fabrication and calculation errors

February 13, 2008, CCP A room cooler leak

February 28, 2008, ECCS voids

March 11, 2008, safety injection tank nitrogen leak

Documents reviewed by the inspectors are listed in the attachment.

The inspectors completed four samples.

b. Findings

An unresolved item (URI) was identified when an operability determination dated

January 22, 2008, was required to ensure that latent fabrication and calculation errors

did not create unacceptable reductions in net positive suction head requirements for

pumps in emergency core cooling systems. This new design information was

associated with the already installed containment recirculation sump strainer

modification. The associated vendor calculation, TD 6002 05, for clean strainer head

loss omitted the head loss component associated with the orifices located in the strainer

support plate. The size of the orifice beneath each strainer tube was not large enough to

prevent head loss in excess of the net positive suction head required per the design

conditions defined in the purchase specification supplied to the strainer vendor. The

additional head loss due to the calculation correction was 2.28 feet. This resulted in

required net positive suction head being less than available. Wolf Creek performed an

operability determination review to demonstrate that the head loss margin could be

recovered. The operability determination on January 22, 2008, addressed the smaller

- 16 - Enclosure

support plate orifice holes by using additional margin gained by separating the head loss

of the RHR and containment spray piping systems to demonstrate lower losses and

additional water inventory in containment prior to containment spray swapover to the

sump. Wolf Creek is replacing the strainer support plate with larger orifices to regain

head loss margin in Refueling Outage 16. However, additional concerns were provided

to the licensee by the vendor on April 1, 2008, addressing nonconservative temperature

correction through the orifices. Subsequent to this, the licensee will need to perform

additional analyses to determine if negative margin existed during the last cycle and if

the new orifice holes will provide positive margin. At the completion of the inspection

period, there were still unresolved questions about the assumptions and results

associated with the calculations used for regaining net positive suction head margin.

These concerns require additional inspection and, when completed, the inspection

results will require significance determination. This issue is considered unresolved

pending additional NRC review of Wolf Creek operability determination calculations:

URI 05000483/2008002-03, Containment Sump Net Positive Suction Head Losses.

1R18 Plant Modifications (71111.18)

.1 Permanent Modification Review

a. The inspectors reviewed key affected parameters associated with energy needs,

materials/replacement components, timing, heat removal, control signals, equipment

protection from hazards, operations, flowpaths, pressure boundary, ventilation boundary,

structural, process medium properties, licensing basis, and failure modes for the one

modification listed below. The inspectors verified that: (1) modification preparation,

staging, and implementation does not impair emergency/abnormal operating procedure

actions, key safety functions, or operator response to loss of key safety functions;

(2) postmodification testing will maintain the plant in a safe configuration during testing

by verifying that unintended system interactions will not occur, SSC performance

characteristics still meet the design basis, the appropriateness of modification design

assumptions, and the modification test acceptance criteria has been met; and (3) the

licensee has identified and implemented appropriate corrective actions associated with

permanent plant modifications

March 6, 2008, containment spray recirculation piping for full flow testing

Documents reviewed by the inspectors are listed in the attachment.

The inspectors completed one sample.

b. Findings

No findings of significance were identified.

.2 Temporary Modification Review

a. Inspection Scope

The inspectors reviewed plant drawings, procedure requirements, and Technical

Specifications to ensure that the below temporary modification was properly

implemented. The inspectors: (1) verified that the modification did not have an affect on

- 17 - Enclosure

system operability/availability, (2) verified that the installation was consistent with the

modification documents, (3) ensured that the post installation test results were

satisfactory and that the impact of the temporary modification on permanently installed

SSCs were supported by the test, (4) verified that the modifications were identified on

control room drawings and that appropriate identification tags were placed on the

affected drawings, and (5) verified that appropriate safety evaluations were completed.

The inspectors verified that licensee identified and implemented any needed corrective

actions associated with temporary modifications.

January 22, 2008, safety injection room cooler temporary modification procedure

February 6, 2008, rod control circuitry monitoring equipment for troubleshooting

Documents reviewed by the inspectors are listed in the attachment.

The inspectors completed two samples.

b. Findings

No findings of significance were identified

1R19 Postmaintenance Testing (71111.19)

a. Inspection Scope

The inspectors selected the below listed postmaintenance test activities of risk

significant systems or components. For each item, the inspectors: (1) reviewed the

applicable licensing basis and/or design-basis documents to determine the safety

functions; (2) evaluated the safety functions that may have been affected by the

maintenance activity; and (3) reviewed the test procedure to ensure it adequately tested

the safety function that may have been affected. The inspectors either witnessed or

reviewed test data to verify that acceptance criteria were met, plant impacts were

evaluated, test equipment was calibrated, procedures were followed, jumpers were

properly controlled, the test data results were complete and accurate, the test equipment

was removed, the system was properly realigned, and deficiencies during testing were

documented. The inspectors also reviewed the USAR and corrective action program to

determine if the licensee identified and corrected problems related to postmaintenance

testing.

January 31, 2008, safety injection Pump A run following planned maintenance

February 15, 2008, EDG B run following planned maintenance

March 5, 2008, centrifugal charging Pump A following planned maintenance

Documents reviewed by the inspectors are listed in the attachment.

The inspectors completed three samples.

- 18 - Enclosure

b. Findings

No findings of significance were identified.

1R20 Outage Activities (71111.20)

.1 Refueling Outage Activities

a. Inspection Scope

The inspectors reviewed the outage safety plan and contingency plans for Wolf Creek

Refueling Outage 16 that started on March 22, 2008, and continued into the next period,

to confirm that the licensee had appropriately considered risk, industry experience, and

previous site-specific problems in developing and implementing a plan that assured

maintenance of defense indepth. During the refueling outage, the inspectors observed

portions of the shutdown and cooldown processes and monitored licensee controls over

the outage activities listed below.

Licensee configuration management, including maintenance of defense indepth

commensurate with the outage safety plan for key safety functions and

compliance with the applicable Technical Specifications when taking equipment

out of service.

Implementation of clearance activities and confirmation that tags were properly

hung and equipment appropriately configured to safely support the work or

testing.

Installation and configuration of reactor coolant pressure, level, and temperature

instruments to provide accurate indication, accounting for instrument error.

Controls over the status and configuration of electrical systems to ensure that

Technical Specification and outage safety plan requirements were met, and

controls over switchyard activities.

Monitoring of decay heat removal processes, systems, and components.

Controls to ensure that outage work was not impacting the ability of the operators

to operate the spent fuel pool cooling system.

Reactor water inventory controls including flow paths, configurations, and

alternative means for inventory addition, and controls to prevent inventory loss.

Controls over activities that could affect reactivity.

Maintenance of secondary containment as required by Technical Specifications.

Refueling activities, including fuel handling and sipping to detect fuel assembly

leakage.

- 19 - Enclosure

Licensee identification and resolution of problems related to refueling outage

activities.

The inspectors will complete this sample in the next quarter.

b. Findings

No findings of significance were identified.

.2 Other Outage Activities

a. Inspection Scope

The inspectors evaluated forced outage activities for an unscheduled outage that began

on January 11, 2008, and continued through January 16, 2008, due to a Technical

Specification required shutdown for voiding in ECCSs. The inspectors reviewed

activities to ensure that the licensee considered risk in developing, planning, and

implementing the outage schedule. The inspectors observed or reviewed the reactor

shutdown and cooldown, outage equipment configuration, risk management, electrical

lineups, selected clearances, control and monitoring of decay heat removal, control of

containment activities, and identification and resolution of problems associated with the

outage. The inspectors observed portions of the reactor startup and heatup.

Documents reviewed by the inspectors are listed in the attachment.

The inspectors completed one sample.

b. Findings

No findings of significance were identified for the forced outage. Aspects of the ECCS

voiding will be contained in NRC Special Inspection Report 05000482/2008-008.

.3 Other Outage Activities

a. Inspection Scope

The inspectors evaluated forced outage activities for an unscheduled outage that began

on March 17, 2008, due to a reactor trip due to XPB03 transformer failure, and continued

through the start of Refueling Outage 16. The inspectors reviewed activities to ensure

that the licensee considered risk in developing, planning, and implementing the outage

schedule. The inspectors observed or reviewed the reactor shut down and cool down,

outage equipment configuration and risk management, electrical lineups, selected

clearances, control and monitoring of decay heat removal, control of containment

activities, and identification and resolution of problems associated with the outage.

Documents reviewed by the inspectors are listed in the attachment.

The inspectors completed one sample.

- 20 - Enclosure

b. Findings

No findings of significance were identified for the forced outage. Additional aspects of

the plant trip are in Section 4OA3.3.

1R22 Surveillance Testing (71111.22)

.1 Routine Surveillance Testing

a. Inspection Scope

The inspectors reviewed the test results for the following activities to determine whether

risk-significant systems and equipment were capable of performing their intended safety

function and to verify testing was conducted in accordance with applicable procedural

and Technical Specification requirements:

January 26, 2008, emergency exhaust system run

February 11-15, 2008, offsite power alignment Technical Specification

surveillance

February 18, 2008, STS BG-002, ECCS vent and void checks

March 6, 2008, EDG A biennial 24-hour endurance and load test

The inspectors observed in-plant activities and reviewed procedures and associated

records to determine whether: any preconditioning occurred; effects of the testing were

adequately addressed by control room personnel or engineers prior to the

commencement of the testing; acceptance criteria were clearly stated, demonstrated

operational readiness, and were consistent with the system design basis; plant

equipment calibration was correct, accurate, and properly documented; as left setpoints

were within required ranges; the calibration frequency was in accordance with Technical

Specifications, the USAR, procedures, and applicable commitments; measuring and test

equipment calibration was current; test equipment was used within the required range

and accuracy; applicable prerequisites described in the test procedures were satisfied;

test frequencies met Technical Specification requirements to demonstrate operability

and reliability; tests were performed in accordance with the test procedures and other

applicable procedures; jumpers and lifted leads were controlled and restored where

used; test data and results were accurate, complete, within limits, and valid; test

equipment was removed after testing; where applicable, test results not meeting

acceptance criteria were addressed with an adequate operability evaluation or the

system or component was declared inoperable; where applicable for safety-related

instrument control surveillance tests, reference setting data were accurately incorporated

in the test procedure; where applicable, actual conditions encountering high resistance

electrical contacts were such that the intended safety function could still be

accomplished; prior procedure changes had not provided an opportunity to identify

problems encountered during the performance of the surveillance or calibration test;

equipment was returned to a position or status required to support the performance of

the safety functions; and all problems identified during the testing were appropriately

documented and dispositioned in the corrective action program.

- 21 - Enclosure

Documents reviewed by the inspectors are listed in the attachment.

The inspectors completed four samples

b. Findings

No findings of significance were identified.

.2 In-service Testing Surveillance

a. Inspection Scope

The inspectors reviewed the test results for the following activities to determine whether

risk-significant systems and equipment were capable of performing their intended safety

function and to verify testing was conducted in accordance with applicable procedural

and Technical Specification requirements:

January 23, 2008, inservice testing of residual heat removal Pump B

February 5, 2008, inservice testing of atmospheric relief Valve D

The inspectors observed in-plant activities and reviewed procedures and associated

records to determine whether: any preconditioning occurred; effects of the testing were

adequately addressed by control room personnel or engineers prior to the

commencement of the testing; acceptance criteria were clearly stated, demonstrated

operational readiness, and were consistent with the system design basis; plant

equipment calibration was correct, accurate, and properly documented; as left setpoints

were within required ranges; and the calibration frequency were in accordance with

Technical Specifications, the USAR, procedures, and applicable commitments;

measuring and test equipment calibration was current; test equipment was used within

the required range and accuracy; applicable prerequisites described in the test

procedures were satisfied; test frequencies met Technical Specification requirements to

demonstrate operability and reliability; tests were performed in accordance with the test

procedures and other applicable procedures; jumpers and lifted leads were controlled

and restored where used; test data and results were accurate, complete, within limits,

and valid; test equipment was removed after testing; where applicable for inservice

testing activities, testing was performed in accordance with the applicable version of

Section XI, American Society of Mechanical Engineers (ASME) Code, and reference

values were consistent with the system design basis; where applicable, test results not

meeting acceptance criteria were addressed with an adequate operability evaluation or

the system or component was declared inoperable; where applicable for safety-related

instrument control surveillance tests, reference setting data were accurately incorporated

in the test procedure; where applicable, actual conditions encountering high resistance

electrical contacts were such that the intended safety function could still be

accomplished; prior procedure changes had not provided an opportunity to identify

problems encountered during the performance of the surveillance or calibration test;

equipment was returned to a position or status required to support the performance of its

safety functions; and all problems identified during the testing were appropriately

documented and dispositioned in the corrective action program. Documents reviewed

are listed in the attachment.

- 22 - Enclosure

Documents reviewed by the inspectors are listed in the attachment.

The inspectors completed two samples.

b. Findings

No findings of significance were identified.

1EP6 Drill Evaluation (71114.06)

a. Inspection Scope

The drill listed below contributed to drill/exercise performance and emergency response

organization performance indicators. The inspectors: (1) observed the training evolution

to identify any weaknesses and deficiencies in classification, notification, and protective

action requirements development activities; (2) compared the identified weaknesses and

deficiencies against licensee identified findings to determine whether the licensee is

properly identifying failures; and (3) determined whether licensee performance is in

accordance with the guidance of the Nuclear Energy Institute (NEI) 99-02 documents

acceptance criteria.

January 31, 2008, loss of all annunciators followed by loss of all offsite power

Documents reviewed by the inspectors are listed in the attachment.

The inspectors completed one sample.

b. Findings

No findings of significance were identified.

2. RADIATION SAFETY

Cornerstone: Occupational Radiation Safety

2OS1 Access Control to Radiologically Significant Areas (71121.01)

Review of Licensee Performance Indicators for the Occupational Exposure Cornerstone

a. Inspection Scope

This area was inspected to assess the licensees performance in implementing physical

and administrative controls for airborne radioactivity areas, radiation areas, high

radiation areas, and worker adherence to these controls. The inspectors used the

requirements in 10 CFR Part 20, the Technical Specifications, and the licensees

procedures required by Technical Specifications as criteria for determining compliance.

During the inspection, the inspectors interviewed the radiation protection manager,

radiation protection supervisors, and radiation workers. The inspectors performed

independent radiation dose rate measurements and reviewed the following items:

Performance indicator events and associated documentation packages reported

by the licensee in the occupational radiation safety cornerstone

- 23 - Enclosure

Controls (surveys, posting, and barricades) of radiation, high radiation, or

airborne radioactivity areas

Radiation work permits, procedures, engineering controls, and air sampler

locations

Conformity of electronic personal dosimeter alarm set points with survey

indications and plant policy; workers knowledge of required actions when their

electronic personnel dosimeter noticeably malfunctions or alarms

Barrier integrity and performance of engineering controls in airborne radioactivity

areas

Physical and programmatic controls for highly activated or contaminated

materials (nonfuel) stored within spent fuel and other storage pools.

Self-assessments, audits, licensee event reports (LERs), and special reports

related to the access control program since the last inspection

Corrective action documents related to access controls

Licensee actions in cases of repetitive deficiencies or significant individual

deficiencies

Radiation work permit briefings and worker instructions

Adequacy of radiological controls, such as required surveys, radiation protection

job coverage, and contamination control during job performance

Dosimetry placement in high radiation work areas with significant dose rate

gradients

Changes in licensee procedural controls of high dose rate - high radiation areas

and very high radiation areas

Controls for special areas that have the potential to become very high radiation

areas during certain plant operations

Posting and locking of entrances to all accessible high dose rate - high radiation

areas and very high radiation areas

Radiation worker and radiation protection technician performance with respect to

radiation protection work requirements

Documents reviewed by the inspector are listed in the attachment.

The inspector completed 20 of the required 21 samples.

- 24 - Enclosure

b. Findings

.1 Introduction. A self-revealing NCV of Technical Specification 5.7.2.a was reviewed

involving the failure to evaluate changing radiological conditions in Room 7604 and

control the area as a locked high radiation area. The violation had very low safety

significance.

Description. On September 26, 2007, mixed bed Demineralizer A was sluiced to the

primary spent resin storage tank. During a followup survey, a localized concentration of

resin was identified in the common sluice header in the 2022 pipe chase of the

radioactive waste building. Operations was unable to perform a lineup to remove the

resin because a procedure for that evolution did not exist. It was decided to leave the

resin in place since it was in a remote location, the area was being controlled as a

locked high radiation area, and it would be removed during the next resin sluice

evolution.

On October 15, 2007, Clearance Order EC-N-005 was developed to vent and drain a

section of piping in preparation for work on Valve ECV0081. The vent path for

Clearance Order EC-N-005 was through Valve ECV0079, located in Room 7604, which

ties into the common resin sluice header. Dose rates in Room 7604 are typically in the

range of 8 - 10 mRem/hour. The clearance order group was not informed of the

localized concentration of resin remaining in the sluice header. In preparation for

hanging the clearance, operations radioactive waste personnel and health physics

personnel discussed the dose rates in the affected areas, valve manipulations, and that

the vent path was hard piped and would not require a temporary hose connection.

On the morning of October 16, 2007, a radwaste person performing routine evolutions

entered Room 7604 and immediately received a dose rate alarm on his electronic

dosimeter. The operator immediately exited the room and notified health physics

personnel. An evaluation of the individuals electronic dosimeter indicated he had

entered an area with a dose rate of 74 mRem/hour. The dosimeter had been set to

alarm at 50 mRem/hour. Radiological surveys of the area taken at 9 a.m. on October 16

indicated dose rates as high as 197 mRem/hour. The area was immediately posted as a

high radiation area. At this time, the licensee did not understand the cause of the

increased radiation levels. Followup surveys were taken at 9 a.m. on October 17 and

indicated that dose rates had increased to 1500 mRem/hour requiring posting and

control as a locked high radiation area. The area was immediately posted and controlled

as a locked high radiation area. Subsequent surveys showed dose rates reached a

maximum of 2500 mRem/hour before a temporary instruction was written to flush the

resin from the common sluice header.

The inspectors determined that health physics personnel failed to perform timely surveys

to identify and control a locked high radiation area. Corrective actions included

immediately posting and controlling the area as a locked high radiation area and

developing a temporary procedure to flush the resin from the common sluice header to

the spent resin storage tank.

Analysis. This finding is more minor because it is associated with the occupational

radiation safety program and process attribute and affected the cornerstone objective, in

that, the failure to properly post and control access to a locked high radiation area has

the potential to increase personnel dose. This occurrence involves the potential for

- 25 - Enclosure

unplanned, unintended dose. Utilizing IMC 0609, Appendix C, Occupational Radiation

Safety Significance Determination Process, the inspector determined that the finding

was of very low safety significance because it did not involve: (1) as low as is

reasonably achievable (ALARA) planning and controls, (2) an overexposure, (3) a

substantial potential for overexposure, or (4) an impaired ability to assess dose. This

finding has a crosscutting aspect in the area of human performance associated with the

work control component because the licensee failed to appropriately plan work activities

by incorporating job site conditions that may impact radiological safety H.3(a).

Enforcement. Technical Specification 5.7.2.a requires that the entryway to high radiation

areas with dose rates greater than 1.0 Rem/hour be conspicuously posted as a high

radiation area and shall be provided with a locked or continuously guarded door or gate.

Contrary to this requirement, the licensee failed to perform a timely survey of Room

7604 and evaluate changing radiological conditions which, required the room to be

posted and controlled as a Locked High Radiation Area. Because the finding is of very

low safety significance and has been entered into the licensees corrective action

program as CR 2007-003934, this violation is being treated as an NCV consistent with

Section VI.A of the Enforcement Policy: NCV 05000482/2008002-04, Failure to Control

Area as a Locked High Radiation Area.

.2 Introduction. The inspectors reviewed a self-revealing NCV of Technical

Specification 5.4.1 for failure to follow a licensee procedure.

Description. On March 29, 2008, while performing radiography at the quality control

vault, a radiographer assistant received a dose rate alarm on his electronic dosimeter.

Radiography evolutions at the site are controlled using a radiation work permit provided

by the health physics department. Radiation Work Permit 08-3021 established a dose

rate alarm setpoint of 500 mRem/hour. The radiography crew properly secured the

radiography source and performed surveys of the source camera. The radiographer and

assistant reviewed the dose received by each individual as indicated on their electronic

dosimeters and, without notifying health physics personnel, decided to continue with

radiography. The alarm condition was noted when the radiographer and assistant

returned to access control to sign off of the radiation work permit. A review of the

dosimeters indicated that the assistant received a total dose of 2.0 mRem with a peak

dose rate of 512 mRem/hour and the radiographer received 2.9 mRem with a peak dose

rate of 476 mRem/hour. Immediate corrective actions included restriction of the

radiographers to log onto the radiation work permit and discussions with the

radiographers and the contractors radiation safety officer. Long-term corrective action is

still being evaluated.

Analysis. This finding is greater than minor because it is associated with the

occupational radiation safety program and process attribute and affected the

cornerstone objective, in that the failure to stop work and notify health physics personnel

for assistance had the potential to increase personnel dose. This occurrence involves

the potential for unplanned, unintended dose. Utilizing IMC 0609, Appendix C,

Occupational Radiation Safety Significance Determination Process, the inspectors

determined that the finding was of very low safety significance because it did not involve:

(1) ALARA planning and controls, (2) an overexposure, (3) a substantial potential for

overexposure, or (4) an impaired ability to assess dose. This finding has a crosscutting

aspect in the area of human performance associated with the decision making

component because the radiographer and assistant failed to contact health physics

- 26 - Enclosure

personnel to discuss the circumstances surrounding the unexpected dose rate alarm

H.1(a).

Enforcement. Technical Specification 5.4.1 requires procedures be established,

implemented, and maintained covering the applicable procedures recommended in

Regulatory Guide 1.33, Appendix A. Section 7 of Appendix A recommends radiation

protection procedures for personnel monitoring. Licensee Procedure AP 25B-100,

Radiation Worker Guidelines, Section 6.2.8 states, in part, If an individuals electronic

dosimeter alarms, the worker shall notify coworkers/health physics and exit the area.

Health physics personnel will then evaluate radiological conditions prior to the

continuation of work. Contrary to this requirement, the radiographer and assistant failed

to notify health physics personnel prior to resuming work activities. Because this failure

to follow a procedure is of very low safety significance and has been entered into the

licensees corrective action program as CR 2008-001181, this violation is being treated

as an NCV, consistent with Section VI.A of the NRC Enforcement Policy: NCV 05000482/2008002-05, Failure to Follow Procedure.

2OS2 ALARA Planning and Controls (71121.02)

Inspection Planning

a. Inspection Scope

The inspectors assessed licensee performance with respect to maintaining individual

and collective radiation exposures ALARA. The inspectors used the requirements in

10 CFR Part 20 and the licensees procedures required by Technical Specifications as

criteria for determining compliance. The inspector interviewed licensee personnel and

reviewed:

Outage or on-line maintenance work activities scheduled during the inspection

period and associated work activity exposure estimates, which were likely to

result in the highest personnel collective exposures

Site-specific ALARA procedures

Integration of ALARA requirements into work procedure and radiation work

permit documents

Workers use of the low-dose waiting areas

Radiation worker and radiation protection technician performance during work

activities in radiation areas, airborne radioactivity areas, or high radiation areas

Self-assessments, audits, and special reports related to the ALARA program

since the last inspection

Resolution through the corrective action process of problems identified through

postjob reviews and postoutage ALARA report critiques

Corrective action documents related to the ALARA program and followup

activities, such as initial problem identification, characterization, and tracking

- 27 - Enclosure

Effectiveness of self-assessment activities with respect to identifying and

addressing repetitive deficiencies or significant individual deficiencies

Documents reviewed by the inspector are listed in the attachment.

The inspector completed 5 of the required 15 samples and 4 of the optional samples.

b. Findings

No findings of significance were identified.

4. OTHER ACTIVITIES

4OA1 Performance Indicator Verification (71151)

.1 Data Submission

a. Inspection Scope

The inspectors performed a review of the data submitted by the licensee for the 4th

quarter 2007, performance indicators for any obvious inconsistencies prior to its public

release in accordance with IMC 0608, Performance Indicator Program.

This review was performed as part of the inspectors normal plant status activities and,

as such, did not constitute a separate inspection sample.

b. Findings

No findings of significance were identified.

.2 Unplanned Scrams per 7000 Critical Hours

a. Inspection Scope

The inspectors sampled licensee submittals for the unplanned scrams per 7000 critical

hours performance indicator for the period from the 4th quarter 2006 through the 4th

quarter 2007. To determine the accuracy of the performance indicator data reported

during those periods, performance indicator definitions and guidance contained in

Revision 5 of the NEI Document 99-02, Regulatory Assessment Performance Indicator

Guideline, were used. The inspectors reviewed the licensees operator narrative logs,

issue reports, event reports and NRC Inspection reports for the period of January 1,

2006, through December 31, 2007, to validate the accuracy of the submittals. The

inspectors also reviewed the licensees issue report database to determine if any

problems had been identified with the performance indicator data collected or

transmitted for this indicator and none were identified

Documents reviewed by the inspectors are listed in the attachment.

The inspectors completed one sample.

- 28 - Enclosure

b. Findings

No findings of significance were identified.

.3 Unplanned Scrams with Complications

a. Inspection Scope

The inspectors sampled licensee submittals for the unplanned scrams with

complications performance indicator for the period from the 4th quarter 2006 through the

4th quarter 2007. To determine the accuracy of the performance indicator data reported

during those periods, performance indicator definitions and guidance contained in

Revision 5 of the NEI Document 99 02, Regulatory Assessment Performance Indicator

Guideline, were used. The inspectors reviewed the licensees operator narrative logs,

issue reports, event reports and NRC Integrated Inspection reports for the period of

January 1, 2006, through December 31, 2007, to validate the accuracy of the submittals.

The inspectors also reviewed the licensees issue report database to determine if any

problems had been identified with the performance indicator data collected or

transmitted for this indicator and none were identified.

Documents reviewed by the inspectors are listed in the attachment.

The inspectors completed one sample.

b. Findings

No findings of significance were identified.

.4 Unplanned Transients per 7000 Critical Hours

a. Inspection Scope

The inspectors sampled licensee submittals for the unplanned transients per

7000 critical hours performance indicator for the period from the 4th quarter 2006

through the 4th quarter 2007. To determine the accuracy of the performance indicator

data reported during those periods, performance indicator definitions and guidance

contained in Revision 5 of the NEI Document 99 02, Regulatory Assessment

Performance Indicator Guidelines, were used. The inspectors reviewed the licensees

operator narrative logs, issue reports, maintenance rule records, event reports and NRC

integrated inspection reports for the period of January 1, 2006, through December 31,

2007, to validate the accuracy of the submittals. The inspectors also reviewed the

licensees issue report database to determine if any problems had been identified with

the performance indicator data collected or transmitted for this indicator and none were

identified.

Documents reviewed by the inspectors are listed in the attachment.

The inspectors completed one sample.

- 29 - Enclosure

b. Findings

No findings of significance were identified.

.5 Occupational Radiation Safety

a. Inspection Scope

The inspectors reviewed licensee documents for occupational exposure control

effectiveness from July 1 through December 31, 2007. The review included corrective

action documentation that identified occurrences in locked high radiation areas (as

defined in the licensees technical specifications), very high radiation areas (as defined

in 10 CFR 20.1003), and unplanned personnel exposures (as defined in NEI 99-02,

"Regulatory Assessment Indicator Guideline," Revision 5). Additional records reviewed

included ALARA records and whole body counts of selected individual exposures. The

inspector interviewed licensee personnel that were accountable for collecting and

evaluating the performance indicator data. In addition, the inspector toured plant areas

to verify that high radiation, locked high radiation, and very high radiation areas were

properly controlled. Performance indicator definitions and guidance contained in

NEI 99-02, Revision 5, were used to verify the basis in reporting for each data element.

Documents reviewed by the inspectors are listed in the attachment.

The inspectors completed one sample.

b. Findings

No findings of significance were identified.

.6 Public Radiation Safety

a. Inspection Scope

The inspectors reviewed licensee documents for Radiological Effluent Technical

specification/Offsite Dose Calculation Manual Radiological Effluent Occurrences from

July 1 through December 31, 2007. Licensee records reviewed included corrective

action documentation that identified occurrences for liquid or gaseous effluent releases

that exceeded performance indicator thresholds and those reported to the NRC. The

inspector interviewed licensee personnel that were accountable for collecting and

evaluating the performance indicator data. Performance indicator definitions and

guidance contained in NEI 99-02, Revision 5, were used to verify the basis in reporting

for each data element.

Documents reviewed by the inspectors are listed in the attachment.

The inspectors completed one sample.

- 30 - Enclosure

b. Findings

No findings of significance were identified

4OA2 Identification and Resolution of Problems (71152)

Cornerstones: Initiating Events, Mitigating Systems, Barrier Integrity, Emergency

Preparedness, Public Radiation Safety, Occupational Radiation Safety, and Physical

Protection

.1 Routine Review of Items Entered Into the Corrective Action Program

a. Inspection Scope

As part of the various baseline inspection procedures discussed in previous sections of

this report, the inspectors routinely reviewed issues during baseline inspection activities

and plant status reviews to verify that they were being entered into the licensees

corrective action program at an appropriate threshold, that adequate attention was being

given to timely corrective actions, and that adverse trends were identified and

addressed. Attributes reviewed included: the complete and accurate identification of the

problem; that timeliness was commensurate with the safety significance; that evaluation

and disposition of performance issues, generic implications, common causes,

contributing factors, root causes, extent of condition reviews, and previous occurrences

reviews were proper and adequate; and that the classification, prioritization, focus, and

timeliness of corrective actions were commensurate with safety and sufficient to prevent

recurrence of the issue. Minor issues entered into the licensees corrective action

program as a result of the inspectors observations are included in the attached list of

documents reviewed.

These routine reviews for the identification and resolution of problems did not constitute

any additional inspection samples. Instead, by procedure they were considered an

integral part of the inspections performed during the quarter and documented in

Section 1 of this report.

b. Findings

No findings of significance were identified.

.2 Daily Corrective Action Program Reviews

a. Inspection Scope

In order to assist with the identification of repetitive equipment failures and specific

human performance issues for follow-up, the inspectors performed a daily screening of

items entered into the licensees corrective action program. This review was

accomplished through inspections of the stations daily condition report packages.

These daily reviews were performed by procedure as part of the inspectors daily plant

status monitoring activities and, as such, did not constitute any separate inspection

samples.

- 31 - Enclosure

b. Findings

No findings of significance were identified.

.3 Selected Issue Follow-up Inspection

a. Inspection Scope

During a review of items entered in the licensees corrective action program, the

inspectors selected the corrective action report listed below for a more indepth review.

The inspectors considered the following during the review of the licensee's actions:

(1) complete and accurate identification of the problem in a timely manner; (2) evaluation

and disposition of operability/reportability issues; (3) consideration of extent of condition,

generic implications, common cause, and previous occurrences; (4) classification and

prioritization of the resolution of the problem; (5) identification of root and contributing

causes of the problem; (6) identification of corrective actions; and (7) completion of

corrective actions in a timely manner.

March 10, 2008, CR 2008-000790, automatic voltage control affected by scaffold

construction

The above constitutes completion of one indepth problem identification and resolution

sample.

Documents reviewed by the inspectors are listed in the attachment.

The inspectors completed one sample.

b. Findings

No findings of significance were identified.

.4 Routine Review of Items Entered into the Corrective Action Program for Access Control

to Radiologically Significant Areas and ALARA Planning and Controls