ML082180851

IR 05000483-08-003, on 3/25 - 6/24/08, Callaway Plant, Operability Evaluations, Postmaintenance Testing, Surveillance Testing, and Identification and Resolution of Problems

ML082180851
Person / Time
Site:	Callaway
Issue date:	08/05/2008
From:	Vincent Gaddy NRC/RGN-IV/DRP/RPB-B
To:	Heflin A Union Electric Co
References
EA-08-190 IR-08-003
Download: ML082180851 (58)

See also: IR 05000483/2008003

Text

UNITED STATES

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

R E GI ON I V

612 EAST LAMAR BLVD , SU I TE 400

AR LI N GTON , TEXAS 76011-4125

August 5, 2008

EA-08-190

Mr. Adam C. Heflin, Senior Vice

President and Chief Nuclear Officer

Union Electric Company

P.O. Box 620

Fulton, MO 65251

SUBJECT: CALLAWAY PLANT - NRC INTEGRATED INSPECTION

REPORT AND NOTICE OF VIOLATION 05000483/2008003

Dear Mr. Heflin:

On June 24, 2008, the U.S. Nuclear Regulatory Commission (NRC) completed an integrated

inspection at your Callaway Plant. The enclosed report documents the inspection results, which

were discussed on June 24, 2008, with you 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, one violation is cited in the enclosed Notice of

Violation (Notice) and the circumstances surrounding this violation are described in detail in the

enclosed report. The violation involved failure to implement corrective actions to preclude the

repetition of void formation in the emergency core cooling piping (EA-08-190). Although

determined to be of very low safety significance (Green), this violation is being cited because

one of the criteria specified in Section VI.A.1 of the NRC Enforcement Policy for a noncited

violation was satisfied. Specifically, AmerenUE failed to restore compliance within a reasonable

time after the violation was last identified in NRC Inspection Report 05000483/2006002-012.

Please note that you are required to respond to this letter and should follow the instructions

specified in the enclosed Notice when preparing your response. The NRC will use your

response, in part, to determine whether further enforcement action is necessary to ensure

compliance with regulatory requirements.

This report also documents four NRC-identified and self-revealing findings of very low safety

significance (Green). These findings were determined to involve violations of NRC

requirements. Additionally, two licensee-identified violations which were determined to be of

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

significance and because they were entered into your corrective action program, the NRC is

treating these findings as NCVs consistent with Section VI.A of the NRC Enforcement Policy. If

you contest these NCVs, you should provide a response within 30 days of the date of this

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

Union Electric Company -2-

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

Administrator, U.S. Nuclear Regulatory Commission Region IV, 612 East Lamar Drive,

Suite 400, Arlington, Texas 76011-4125; the Director, Office of Enforcement, U.S. Nuclear

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

Callaway Plant.

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

enclosures 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,

Projects Branch B

Division of Reactor Projects

Docket: 50-483

License: NPF-30

Enclosures: Notice of Violation and

NRC Inspection Report 05000483/2008003

w/attachment: Supplemental Information

cc w/enclosure: Rick A. Muench, President and

John ONeill, Esq. Chief Executive Officer

Pillsbury Winthrop Shaw Pittman LLP Wolf Creek Nuclear Operating Corporation

2300 N. Street, N.W. P.O. Box 411

Washington, DC 20037 Burlington, KS 66839

Scott A. Maglio, Assistant Manager Kathleen Smith, Executive Director and

Regulatory Affairs Kay Drey, Representative Board of

AmerenUE Directors

P.O. Box 620 Missouri Coalition for the Environment

Fulton, MO 65251 6267 Delmar Boulevard, Suite 2E

St. Louis City, MO 63130

Missouri Public Service Commission

Governors Office Building Lee Fritz, Presiding Commissioner

200 Madison Street Callaway County Courthouse

P.O. Box 360 10 East Fifth Street

Jefferson City, MO 65102-0360 Fulton, MO 65251

H. Floyd Gilzow Les H. Kanuckel, Manager

Deputy Director for Policy Quality Assurance

Missouri Department of Natural Resources AmerenUE

P. O. Box 176 P.O. Box 620

Jefferson City, MO 65102-0176 Fulton, MO 65251

Union Electric Company -3-

Director, Missouri State Emergency Certrec Corporation

Management Agency 4200 South Hulen, Suite 422

P.O. Box 116 Fort Worth, TX 76109

Jefferson City, MO 65102-0116

Keith G. Henke, Planner III

Scott Clardy, Director Division of Community and Public Health

Section for Environmental Public Health Office of Emergency Coordination

Missouri Department of Health and Missouri Department of Health and

Senior Services Senior Services

P.O. Box 570 930 Wildwood,

Jefferson City, MO 65102-0570 P.O. Box 570

Jefferson City, MO 65102

Luke H. Graessle, Manager

Regulatory Affairs Technical Services Branch Chief

AmerenUE FEMA Region VII

P.O. Box 620 2323 Grand Boulevard, Suite 900

Fulton, MO 65251 Kansas City, MO 64108-2670

Thomas B. Elwood, Supervising Engineer Ronald L. McCabe, Chief

Regulatory Affairs and Licensing Technological Hazards Branch

AmerenUE National Preparedness Division

P.O. Box 620 DHS/FEMA

Fulton, MO 65251 9221 Ward Parkway, Suite 300

Kansas City, MO 64114-3372

Union Electric Company -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 (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)

M. Cox, OEDO RIV Coordinator (Mark.Cox@nrc.gov)

OEMail.Resource@nrc.gov

Enforcement Officer (Michael.Vasquez@nrc.gov)

Chief Allegation Coordination/Enforcement Staff (William.Jones@nrc.gov)

Office of Enforcement (Alexander.Sapountizis@nrc.gov)

ROPreports

CWY Site Secretary (Dawn.Yancey@nrc.gov)

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

Publicly Available Non-Publicly Available Sensitive ;Non-Sensitive

R:\_Reactors\_CW\2008\CW 2008003RP-DED.doc ML 082180851

RIV:SRI:DRP/B C:DRS/OB C:DRS/PSB1 C:DRS/EB2 C:DRS/EB1

DDumbacher RELantz MPShannon NFO'Keefe RLBywater

/RA/ VGGaddy for /RA/ /RA/ /RA/ MFRunyan for /RA/

07/29/2008 07/9/2008 07/14/2008 07/15/2008 07/11/2008

C:DRS/PSB2 DRS/SRA ACES C:DRP/B D:DRP

GEWerner DPLoveless GMVasquez VGGaddy DDChamberlain

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

07/17/2008 07/15/2008 07/24/2008 08/5/2008 07/28/2008

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

NOTICE OF VIOLATION

AmerenUE Docket 50-483

Callaway Plant License NPF-30

EA-08-190

During an NRC inspection conducted March 24 through June 24, 2008, a violation of NRC

requirements was identified. In accordance with the NRC Enforcement Policy, the violation is

listed below:

10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, requires, in part, that

measures shall be established to ensure that, for significant conditions adverse to

quality, the cause of the condition is determined and corrective action taken to preclude

repetition.

Contrary to this, from December 26, 2006, through May 21, 2008, the licensee failed to

take corrective actions to preclude repetition of safety-related emergency core cooling

system pipe voiding, and the licensee determined that this condition was a significant

condition adverse to quality.

This violation is associated with a Green Significance Determination Process finding.

Pursuant to the provisions of 10 CFR 2.201, AmerenUE is hereby required to submit a written

statement or explanation to the U.S. Nuclear Regulatory Commission, ATTN: Document

Control Desk, Washington, DC 20555 with a copy to the Regional Administrator, Region IV,

and a copy to the NRC Senior Resident Inspector at the facility that is the subject of this Notice

of Violation (Notice), within 30 days of the date of the letter transmitting this Notice. This reply

should be clearly marked as a "Reply to Notice of Violation EA-08-190," and should include:

(1) the reason for the violation, or, if contested, the basis for disputing the violation or severity

level, (2) the corrective steps that have been taken and the results achieved, (3) the corrective

steps that will be taken to avoid further violations, and (4) the date when full compliance will be

achieved. Your response may reference or include previous docketed correspondence, if the

correspondence adequately addresses the required response. If an adequate reply is not

received within the time specified in this Notice, an order or a Demand for Information may be

issued as to why the license should not be modified, suspended, or revoked, or why such other

action as may be proper should not be taken. Where good cause is shown, consideration will

be given to extending the response time.

If you contest this enforcement action, you should also provide a copy of your response, with

the basis for your denial, to the Director, Office of Enforcement, United States Nuclear

Regulatory Commission, Washington, DC 20555-0001.

Because your response will be made available electronically for public inspection in the NRC

Public Document Room or from the NRCs document system (ADAMS), accessible from the

NRC Web site at http://www.nrc.gov/reading-rm/adams.html, to the extent possible, it should not

include any personal privacy, proprietary, or safeguards information so that it can be made

available to the public without redaction. If personal privacy or proprietary information is

necessary to provide an acceptable response, then please provide a bracketed copy of your

response that identifies the information that should be protected and a redacted copy of your

response that deletes such information. If you request withholding of such material, you must

specifically identify the portions of your response that you seek to have withheld and provide in

-1- Enclosure 1

detail the bases for your claim of withholding (e.g., explain why the disclosure of information will

create an unwarranted invasion of personal privacy or provide the information required by

10 CFR 2.390(b) to support a request for withholding confidential commercial or financial

information). If safeguards information is necessary to provide an acceptable response, please

provide the level of protection described in 10 CFR 73.21.

Dated this 5th day of July 2008

-2- Enclosure 1

U.S. NUCLEAR REGULATORY COMMISSION

REGION IV

Docket: 50-483

License: NPF-30

Report: 05000483/2008003

Licensee: Union Electric Company

Facility: Callaway Plant

Location: Junction Highway CC and Highway O

Fulton, MO

Dates: March 25 - June 24, 2008

Inspectors: D. Dumbacher, Senior Resident Inspector

J. Groom, Resident Inspector

J. Drake, Senior Reactor Inspector, Plant Support, Branch 2

G. Guerra, CHP, Health Physicist, Plant Support Branch 1

Approved By: V. Gaddy, Chief, Project Branch B

Division of Reactor Projects

-1- Enclosure 2

SUMMARY OF FINDINGS

IR 05000483/2008003: 3/25 - 6/24/2008; Callaway Plant, Operability Evaluations,

Postmaintenance Testing, Surveillance Testing, and Identification and Resolution of Problems.

This report covered a 3-month period of inspection by resident inspectors. 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 assigned a severity level after NRC

management 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 3.5.2, "Emergency Core Cooling Systems," after an inadequate

surveillance procedure resulted in the licensee failing to maintain the emergency

core cooling system full of water as required per Technical Specification 3.5.2.

On May 21, 2008, Callaway Plant engineering discovered that a section of the

cold leg recirculation piping, specifically the discharge of the residual heat

removal pumps to the safety injection pumps, contained 6.6 cubic feet of air.

Callaway monthly surveillance Procedure OSP-SA-00003, "Emergency Core

Cooling Flow Path Verification and Venting," had a purpose to: "Verify the ECCS

is full of water," in accordance with Technical Specification Surveillance

Requirement 3.5.2.3. The monthly verification and vent procedure was not

comprehensive enough to ensure all the emergency core cooling system was full

of water.

This finding was more than minor because it was similar to Example 3e of NRC

Inspection Manual Chapter 0612, Appendix E, "Examples of Minor Issues," and

met the Not Minor If, criteria because the failure to meet the licensees

administrative requirement for allowable void fraction impacted the ability of the

Train A safety injection system to function upon initiation of high-pressure

recirculation. This finding affected the mitigating systems cornerstone procedure

quality attribute. Using the Manual Chapter 0609.04, Phase 1 - Initial Screening

and Characterization of Findings, the inspectors determined that this finding

should be evaluated using the Phase 2 process described in Manual

Chapter 0609, Appendix A, Determining the Significance of Reactor Inspection

Findings for At-Power Situations. As described in Section III, of Appendix A,

given that the presolved table did not contain a suitable target or surrogate for

this finding, the senior reactor analyst used the risk-informed notebook to

evaluate the significance of this finding affecting only high-pressure recirculation

as very low risk significance (Green). This finding has a crosscutting aspect in

the area of human performance associated with the decision making component

because the licensee failed to use conservative assumptions in decision making

and did not adopt a requirement to demonstrate that a single vent valve was

sufficient to vent the affected line rather than assuming that an additional

-2- Enclosure 2

installed valve was not necessary to completely fill, vent, and test the line H.1(b)

(Section 1R15).

• Green. A self-revealing noncited violation of 10 CFR Part 50, Appendix B,

Criterion XVI, "Corrective Action," was identified after the licensee failed to

promptly correct leakage from diesel generator jacket water o-rings. On

February 20, 2008, during a normal surveillance run of Emergency Diesel

Generator B, Callaway operations personnel identified an approximately

80 drop-per-minute jacket water leak caused by premature failure of Nitrile type

o-rings. Following restoration of Emergency Diesel Generator B, the licensee

re-evaluated the preventative maintenance frequency for jacket water o-ring

replacement and reduced the replacement frequency from once every 3 years to

once every refueling cycle. Then, on May 28, 2008, during a routine surveillance

run of Emergency Diesel Generator A, Callaway operations personnel identified

that Emergency Diesel Generator A had a 200 drop-per-minute jacket water leak.

Similar to the condition observed on Emergency Diesel Generator B on

February 20, 2008, the source of the leakage was from Nitrile type o-rings within

the jacket water system. The o-rings responsible for jacket water leakage were

found to be of similar age to those that failed during the February 20, 2008,

surveillance but had not been replaced despite the change to the licensee's

preventive maintenance frequency.

This finding, failure to implement adequate corrective actions for degraded Nitrile

type o-rings in Emergency Diesel Generator A after previously identifying the

adverse condition on Emergency Diesel Generator B, was more than minor

because, if left uncorrected, degraded diesel generator jacket water o-rings could

become a more significant safety concern. This finding affected the mitigating

systems cornerstone. Using Manual Chapter 0609.04, Phase 1 - Initial

Screening and Characterization of Findings, this finding was determined to be of

very low safety significance because it was a design deficiency confirmed not to

result in loss of operability. This finding has a crosscutting aspect in the area of

human performance associated with the work controls component because the

licensee failed to plan work activities to support long-term equipment reliability by

addressing known degraded conditions in a more reactive than preventative

manner H.3(b) (Section 1R19).

• Green. The inspectors identified a violation of 10 CFR Part 50, Appendix B,

Criterion XVI, "Corrective Action," because the licensee failed to take corrective

actions to preclude repetition of void formation in emergency core cooling system

piping, a significant condition adverse to quality. After experiencing void

formations in 2005 and 2006, the NRC identified violations of Criterion XVI.

However, licensee corrective actions did not preclude repetition of void

formations that were discovered on May 21, 2008. On that date, Callaway Plant

engineering performed ultrasonic inspection of the safety injection system

common suction piping Line EM023-HCB - 6" and discovered a 6.6 cubic foot

voided area. This exceeded the allowable void fraction of 2.1 cubic feet required

for operability. This voided piping, determined to have existed for over a year,

was caused by relief valve maintenance on Valve EM8858A (May 7, 2007). The

maintenance restoration failed to perform an adequate fill and vent to ensure the

suction pipe was full of water. The inspectors identified several related examples

where the licensee had performed either inadequate operating experience

-3- Enclosure 2

evaluations, inadequate extent of condition reviews, or inadequate procedure

corrections. The violation is being cited in a Notice of Violation because the

licensee failed to restore compliance with a reasonable time after a violation was

last identified in 2006.

This finding, failure to restore compliance to prevent recurrence of emergency

core cooling system voids, was more than minor because it is similar to

Example 3e of NRC Inspection Manual Chapter 0612, Appendix E, "Examples of

Minor Issues," criteria because the failure impacted the ability of the emergency

core cooling system to function upon initiation of high-pressure recirculation.

Using the Manual Chapter 0609.04, Phase 1 - Initial Screening and

Characterization of Findings, the inspectors determined that this finding should

be evaluated using the Phase 2 process described in Manual Chapter 0609,

Appendix A, Determining the Significance of Reactor Inspection Findings for

At-Power Situations. As described in Section III, of Appendix A, given that the

presolved table did not contain a suitable target or surrogate for this finding, the

senior reactor analyst used the risk-informed notebook to evaluate the

significance of this finding as very low risk significance (Green). This finding has

a crosscutting aspect in the area of problem identification and resolution

associated with the corrective action program component because AmerenUE

failed to thoroughly evaluate voiding problems such that the resolutions

addressed causes and extent of condition, as necessary P.1(c) (Section 4OA2).

Cornerstone: Barrier Integrity

• Green. A self-revealing noncited violation of 10 CFR Part 50, Appendix B,

Criterion III, Design Control, was identified after determining that the licensee

had not adequately selected and reviewed the suitability of the design of the

containment air cooler control circuitry. On March 26, 2008, Containment Air

Cooler A fan shut down when shifted from fast to slow speed. Troubleshooting

by the licensee determined that voltage was lost to the control power circuitry

when the fast speed thermal overload tripped. Since the overload contacts were

wired in series, Containment Air Cooler A experienced a complete loss of control

power rendering it inoperable. The licensee determined the trip to be caused by

operation of containment air coolers in fast speed, during a period of higher than

normal containment pressure. The licensee analyzed the potential impact of the

newly discovered adverse containment cooler design vulnerability against design

basis accident scenarios. The licensee determined that a hot zero power main

steam line break results in a delayed safety injection signal allowing the fan

motor overloads to trip prior to being shed by the load sequencer. The

containment air coolers would then experience a complete loss of control power

and would not be capable of automatically restarting in slow speed. The analysis

revealed that the peak containment pressure limit of 48.1 psig would be

preserved. The licensee submitted a licensee event report as required by

10 CFR 50.73 since the inadequate containment air cooler control circuitry

resulted in a condition prohibited by the plants Technical Specifications.

This finding, failure to ensure the design of the containment air cooler control

circuitry was suitable for all plant conditions, was more than minor because it was

associated with the barrier integrity cornerstone attribute of design control and

affects the associated cornerstone objective to provide reasonable assurance

-4- Enclosure 2

that physical design barriers protect the public from radio nuclide releases

caused by accidents or releases. Using Manual Chapter 0609, Appendix H,

Containment Integrity Significance Determination Process," this finding was

determined to be a Type B finding since it was related to a degraded condition

that has potentially important implications for the integrity of the containment,

without affecting the likelihood of core damage. This finding was found to be of

very low safety significance because containment coolers are structures,

systems or components that are not significant contributors to the large early

release frequency. The inspectors determined that this finding does not have a

crosscutting aspect associated with it since the performance deficiency was not

indicative of current licensee performance (Section 1R15).

• Green. The inspectors identified a noncited violation of Technical

Specification 5.4.1.a, Procedures, after Callaway control room operators

improperly entered a wrong Technical Specification action statement due to the

failure to maintain the Technical Specification Bases current. On June 17, 2008,

during surveillance testing, Valve EMHV8823 failed to indicate fully closed.

Since EMHV8823 is an isolation valve for containment Penetration 49, the

licensee entered Technical Specification 3.6.3, Containment Isolation Valves,

Condition C, with an action to restore the valve to an operable status or isolate

the penetration within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. Approximately 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> after Valve EMHV8823

had been declared inoperable, Callaway licensing personnel contacted the

control room and informed them of an approved Technical Specification Bases

change that did not allow Technical Specification 3.6.3, Condition C, to be

applicable to containment Penetration 49. The Technical Specification Bases

change was effective May 1, 2008, but had not been issued to the control room.

The licensee determined that the more restrictive Technical Specification 3.6.3,

Condition A, should have been entered with an action to isolate the affected

penetration within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. The licensee performed a containment entry

following discovery of entry into Technical Specification 3.6.3, Condition A, and

found that Valve EMHV8823 failed its surveillance due to out of adjustment

position indicator limit switches. The valve was verified closed and isolated

allowing exit from Technical Specification 3.6.3, Condition A.

This finding, failure to ensure the Technical Specification Bases were maintained

current and available to the Callaway control room staff, was more than minor

because if left uncorrected, the failure to maintain the Technical Specification

Bases current could become a more significant safety concern. This finding was

determined to affect the barrier integrity cornerstone. Using Manual

Chapter 0609.04, Phase 1 - Initial Screening and Characterization of Findings,

this finding is determined to be of very low safety significance since this finding

did not represent an actual open pathway in the physical integrity of reactor

containment and did not involve an actual reduction in function of hydrogen

ignitors in the reactor containment. This finding has a crosscutting aspect in the

area of human performance associated with the decision making component

because the licensee failed to communicate, in a timely manner, decisions to

personnel who have a need to know the information in order to perform work

safely H.1(c) (Section 1R22).

-5- Enclosure 2

B. Licensee-Identified Violations

Two violations of very low safety significance, which were identified by the licensee,

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

licensee have been entered into the licensees corrective action program. These

violations and corrective action tracking numbers are listed in Section 4OA7.

-6- Enclosure 2

REPORT DETAILS

Summary of Plant Status

AmerenUE operated the Callaway Plant at near 100 percent for the entire quarter.

1. REACTOR SAFETY

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

Emergency Preparedness

1R01 Adverse Weather Protection (71111.01)

.1 Readiness of Offsite and Alternate AC Power System

a. Inspection Scope

The inspectors reviewed the licensees plant features, training lesson plans, and

procedures for operation and continued availability of offsite and alternate AC power

systems to verify they were appropriate. The review included communication protocols

and agreement procedures between the transmission system operator and the nuclear

power plant to verify that appropriate information is exchanged when issues arise that

could impact the offsite power system. Specifically, the procedures were verified to

ensure they specified:

• Required actions needed when notified by the transmission system operator that

posttrip voltage of the offsite power system would not be acceptable to assure

the continued operation of safety related loads without transferring to the onsite

power supply.

• Compensatory actions needed when it is not possible to predict the posttrip

voltage at the nuclear power plant for current grid conditions.

• Required assessment of plant risk based on maintenance activities which could

affect grid reliability, or the ability of the transmission system to provide the offsite

power system.

• Required communications between the nuclear power plant and the transmission

system operator when changes at the nuclear power plant could impact the

transmission system, or when the capability of the transmission system to

provide adequate offsite system power is challenged.

On May 16, 2008, the inspectors evaluated the licensee staffs preparations for summer

readiness of offsite and AC power systems against the sites procedures and determined

that the staffs actions were adequate. Documents reviewed are listed in the

attachment.

These activities constituted one readiness of offsite power inspection sample as defined

by Inspection Procedure 71111.01.

-7- Enclosure 2

b. Findings

No findings of significance were identified.

.2 Readiness for Impending Adverse Weather Conditions

a. Inspection Scope

On May 2, 2008, the inspectors completed a review of the licensee's readiness for

impending adverse weather involving severe thunderstorms. The inspectors:

(1) reviewed plant procedures, the Final Safety Analysis Report (FSAR), 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

emergency diesel generators and offsite power systems to ensure that adverse weather

protection features were sufficient to support operability; (3) reviewed maintenance

records to determine that applicable surveillance requirements were current before the

anticipated severe thunderstorms developed; and (4) reviewed plant modifications,

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

challenged plant operation. Documents reviewed by the inspectors are listed in the

attachment.

These activities constituted one readiness for impending adverse weather inspection

sample as defined by Inspection Procedure 71111.01.

b. Findings

No findings of significance were identified.

1R04 Equipment Alignments (71111.04)

.1 Quarterly Partial System Walkdowns

a. Inspection Scope

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

systems:

• June 3, 2008, Train A auxiliary feedwater system while the Train B motor-driven

auxiliary feedwater pump was out of service for planned maintenance.

• June 10, 2008, Train A 480 volt NG Class 1E switchgear while the Train B

emergency diesel generator was out of service for planned and emergent

maintenance issues.

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 discrepancies that could impact the function of the system, and, therefore,

potentially increase risk. The inspectors reviewed applicable operating procedures,

system diagrams, FSAR, Technical Specification requirements, outstanding work orders,

corrective action documents, and the impact of ongoing work activities on redundant

trains of equipment in order to identify conditions that could have rendered the systems

-8- Enclosure 2

incapable of performing their intended functions. The inspectors also walked down

accessible portions of the systems to verify components and support equipment were

aligned correctly and were 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. Documents

reviewed are listed in the attachment.

These activities constituted two partial system walkdown samples as defined by

Inspection Procedure 71111.04.

b. Findings

No findings of significance were identified.

.2 Complete System Walkdown (71111.04S)

a. Inspection Scope

On April 17, 2008, the inspectors performed a complete system alignment inspection of

Train B of the residual heat removal system to verify the functional capability of the

system. The inspectors selected this system because it was considered both

safety-significant and risk-significant in the licensees probabilistic risk assessment. The

inspectors walked down the system to review mechanical and electrical equipment line

ups, electrical power availability, system pressure and temperature indications, as

appropriate, component labeling, component lubrication, component and equipment

cooling, hangers and supports, operability of support systems, and to ensure that

ancillary equipment or debris did not interfere with equipment operation. The inspectors

reviewed a sample of past and outstanding work orders to determine whether any

deficiencies significantly affected the system function. In addition, the inspectors

reviewed the corrective action program database to ensure that system equipment

alignment problems were being identified and appropriately resolved. The documents

used for the walkdown and issue review are listed in the attachment.

These activities constituted one complete system walkdown sample as defined by

Inspection Procedure 71111.04.

b. Findings

No findings of significance were identified.

-9- Enclosure 2

1R05 Fire Protection (71111.05)

.1 Quarterly Fire 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:

• March 27, 2008, Fire Area C-21, Lower Cable Spreading Room

• April 16, 2008, Fire Area A-17, Electrical Penetration Room (South)

• April 25, 2008, Condensate Storage Tank

• April 29, 2008, Fire Area A-23, Main Steam and Feedwater Isolation Valve

Enclosure

• April 30, 2008, Reactor Building

• June 18, 2008, Fire Area A-1, North Pipe Chase

The inspectors reviewed areas to assess if the licensee 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 implemented adequate compensatory

measures for out of service, degraded or inoperable fire protection equipment, systems,

or features in accordance with the licensees fire plan. The inspectors selected fire

areas based on their overall contribution to internal fire risk as documented in the plants

Individual Plant Examination of External Events with later additional insights, their

potential to impact equipment which could initiate or mitigate a plant transient, or their

impact on the plants ability to respond to a security event. 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. Documents reviewed are listed in the

attachment.

These activities constituted six quarterly fire protection inspection samples as defined by

Inspection Procedure 71111.05.

b. Findings

No findings of significance were identified.

.2 Annual Fire Protection Drill Observation (71111.05A)

a. Inspection Scope

On March 27, 2008, the inspectors observed a fire brigade activation due to a report of

smoke in the laundry decontamination area. The observation evaluated the readiness of

- 10 - Enclosure 2

the plant fire brigade to fight fires. The inspectors verified that the licensee staff

identified deficiencies; openly discussed them in a self-critical manner at the drill debrief,

and took appropriate corrective actions. Specific attributes evaluated were: (1) proper

wearing of turnout gear and self-contained breathing apparatus; (2) proper use and

layout of fire hoses; (3) employment of appropriate fire fighting techniques; (4) sufficient

firefighting equipment brought to the scene; (5) effectiveness of fire brigade leader

communications, command, and control; (6) search for victims and propagation of the

fire into other plant areas; (7) smoke removal operations; (8) utilization of preplanned

strategies; (9) adherence to the preplanned drill scenario; and (10) drill objectives.

Documents reviewed are listed in the attachment.

These activities constituted one annual fire protection inspection sample as defined by

Inspection Procedure 71111.05.

b. Findings

No findings of significance were identified.

1R06 Flood Protection Measures (71111.06)

Internal Flooding

a. Inspection Scope

The inspectors reviewed selected risk-significant plant design features and licensee

procedures intended to protect the plant and its safety related equipment from internal

flooding events. The inspectors reviewed flood analyses and design documents,

including the FSAR, engineering calculations, and abnormal operating procedures for

licensee commitments. The inspectors reviewed licensee drawings to identify areas and

equipment that may be affected by internal flooding caused by the failure or

misalignment of nearby sources of water. The inspectors also reviewed the licensees

corrective actions for previously identified flood-related items. The inspectors performed

a walkdown of the following plant area to assess the adequacy of any watertight doors

and verify drains and sumps were clear of debris and operable, and that the licensee

complied with its flooding related commitments:

• June 23, 2008, Control Building West Corridor

The document reviewed during this inspection is listed as follows:

• Callaway Action Request 200805189

This inspection constituted one internal flooding sample as defined in Inspection

Procedure 71111.06.

b. Findings

No findings of significance were identified.

- 11 - Enclosure 2

1R11 Licensed Operator Requalification Program (71111.11)

a. Inspection Scope

On June 2, 2008, the inspectors observed a crew of licensed operators perform a

Cycle 08-3 as found scenario in the plants simulator to verify that operator performance

was adequate, evaluators were identifying and documenting crew performance

problems, and that training was being conducted in accordance with licensee

procedures. The scenario involved an operating design basis earthquake with a lockout

on essential 4 kV Bus NB01. The inspectors evaluated the crew in the following areas:

• Licensed operator performance

• Crew clarity and formality of communications

• Ability to take timely actions in the conservative direction

• Prioritization, interpretation, and verification of annunciator alarms

• Correct use and implementation of abnormal and emergency procedures

• Control board manipulations

• Oversight and direction from supervisors

• Ability to identify and implement appropriate Technical Specification actions and

Emergency Plan actions and notifications

The crews performance in these areas was compared to pre-established operator action

expectations and successful critical task completion requirements. Documents reviewed

are listed in the attachment.

This inspection constituted one quarterly licensed operator requalification program

sample as defined in Inspection Procedure 71111.11.

b. Findings

No findings of significance were identified.

1R12 Maintenance Effectiveness (71111.12)

a. Inspection Scope

The inspectors evaluated degraded performance issues involving the following

risk-significant systems:

• May 15, 2008, Callaway Action Request (CAR) 200801644, an additional anode

was found in the north end of the Train A emergency diesel generator intercooler

• May 15, 2008, CAR 200802854, KKJ01A (Train A emergency diesel generator)

engine oil sump high

- 12 - Enclosure 2

The inspectors reviewed events such as where ineffective equipment maintenance has

resulted in valid or invalid automatic actuations of risk-important systems and

independently verified the licensee's actions to address system performance or condition

problems in terms of the following:

• Implementing appropriate work practices

• Identifying and addressing common cause failures

• Scoping of systems in accordance with 10 CFR 50.65(b) of the maintenance rule

• Characterizing system reliability issues for performance

• Charging unavailability time

• Trending key parameters for condition monitoring

• Ensuring 10 CFR 50.65(a)(1) or (a)(2) classification or reclassification

• Verifying appropriate performance criteria for structures, systems, and

components/functions classified as (a)(2) or appropriate and adequate goals and

corrective actions for systems classified as (a)(1)

The inspectors assessed performance issues with respect to the reliability, availability,

and condition monitoring of the system. The inspectors verified maintenance

effectiveness issues were entered into the corrective action program with the appropriate

significance characterization. Documents reviewed are listed in the attachment.

This inspection constituted two quarterly maintenance effectiveness samples as defined

in Inspection Procedure 71111.12Q.

b. Findings

No findings of significance were identified.

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

a. Inspection Scope

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:

• April 3, 2008, Routine - Work on turbine-driven auxiliary feedwater

Valve KAPCV-0102

• April 21, 2008, Emergency Diesel Generator A lube oil trouble shooting

• April 28, 2008, Routine - associated with Loose Creek-Callaway 345 kV line

outage

- 13 - Enclosure 2

• June 10, 2008, Risk management actions associated with Emergency Diesel

Generator B jacket water o-ring replacement outage

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,

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

requirements were met. Documents reviewed are listed in the attachment.

These activities constituted four samples as defined by Inspection Procedure 71111.13.

b. Findings

No findings of significance were identified.

1R15 Operability Evaluations (71111.15)

a. Inspection Scope

The inspectors reviewed the following issues:

• March 26, 2008, CARs 200802348, 200802365, and 200802264, Containment

coolers inoperable in fast speed

• April 4, 2008, CARs 200800461 and 200802625, Containment recirculation sump

operability determination, Revisions 3 and 4

• April 9, 2008, Source Range Channel N32 inoperable due to a failed surveillance

• April 23, 2008, Component cooling water system following Valve EGHV0069

failing inservice test stroke time surveillance

• April 30, 2008, CAR 200803465, Emergency diesel generator Garlock flexible

expansion joints

• May 6, 2008, CAR 200803462, Voiding identified in containment spray pump

piping from sump

• May 22, 2008, CAR 200904000, Line EM-023 allowable void fraction exceeded

The inspectors selected potential operability issues based on the risk significance of the

associated components and systems. The inspectors evaluated the technical adequacy

of the evaluations to ensure that Technical Specification operability was properly justified

and the subject component or system remained available such that no unrecognized

increase in risk occurred. The inspectors compared the operability and design criteria in

the appropriate sections of the Technical Specifications and FSAR to the licensees

- 14 - Enclosure 2

evaluations to determine whether the components or systems were operable. Where

compensatory measures were required to maintain operability, the inspectors

determined whether the measures in place would function as intended and were

properly controlled. The inspectors determined, where appropriate, compliance with

bounding limitations associated with the evaluations. Additionally, the inspectors

reviewed a sample of corrective action documents to verify that the licensee was

identifying and correcting deficiencies associated with operability evaluations.

Documents reviewed are listed in the attachment.

This inspection constituted seven samples as defined in Inspection Procedure 71111.15.

b. Findings

.1 Introduction. A self-revealing Green noncited violation (NCV) of 10 CFR Part 50,

Appendix B, Criterion III, Design Control, was identified after determining that the

licensee had not adequately selected and reviewed the suitability of the design of the

containment air cooler control circuitry.

Description. On March 26, 2008, Containment Air Cooler A fan shut down when shifted

from fast to slow speed. Troubleshooting by the licensee determined that voltage was

lost to the control power circuitry when the fast speed thermal overload tripped. Since

the overload contacts were wired in series, Containment Air Cooler A experienced a

complete loss of control power rendering it inoperable. AmerenUE personnel noted that

Precaution 3.6 of Procedure OTN-GN-00001, Containment Cooling and CRDM

Cooling, Revision 14, cautioned that high pressure and cool temperatures across

containment coolers will cause the coolers to operate close to the setpoint of the thermal

overloads. However, the licensees operability determination dismissed the 1987

precaution as not having a technical basis believing it was implemented to address

discrepancies in motor overload setpoints. Later, the licensee determined that operation

of containment air coolers in fast speed, during a period of higher than normal

containment pressure, challenged the fast speed thermal overload setpoint and resulted

in the trip of Containment Air Cooler A on March 26, 2008. As an interim measure to

prevent a trip from fast speed, the licensee imposed a standing order to maintain the

containment coolers in slow speed.

The licensee analyzed the potential impact of the newly discovered adverse containment

cooler design vulnerability against design basis accident scenarios. The licensee

determined that a hot zero power main steam line break results in a delayed safety

injection signal allowing the fan motor overloads to trip prior to being shed by the load

sequencer. The containment air coolers would then experience a complete loss of

control power and would not be capable of automatically restarting in slow speed. The

analysis revealed that in this scenario, utilizing assumed accident conditions, the peak

containment pressure would exceed the 48.1 psig limit described in the FSAR.

However, analysis using actual plant conditions determined that the peak containment

pressure limit of 48.1 psig would be preserved. The licensee submitted a licensee event

report (LER) as required by 10 CFR 50.73 since the inadequate containment air cooler

control circuitry resulted in a condition prohibited by the plants Technical Specifications.

The inspectors review of the licensees LER is described in Section 4OA3 of this report.

To address the design deficiency associated with the containment air cooler control

circuitry, the licensee completed a modification in April 2008 to reconfigure the circuit

- 15 - Enclosure 2

such that tripping of the fast speed overloads would not impact the safety-related slow

speed function of the containment air coolers.

Analysis. The performance deficiency associated with this finding involved the

licensees failure to ensure the design of the containment air cooler control circuitry was

suitable for all plant conditions. This finding was greater than minor because it was

associated with the barrier integrity cornerstone attribute of design control and affects

the associated cornerstone objective to provide reasonable assurance that physical

design barriers protect the public from radio nuclide releases caused by accidents or

releases. Using Manual Chapter 0609, Appendix H, Containment Integrity Significance

Determination Process," this finding was determined to be a Type B finding since it was

related to a degraded condition that has potentially important implications for the integrity

of the containment, without affecting the likelihood of core damage. This finding was

found to be of very low safety significance since containment coolers are structures,

systems, and components that have no impact on large early release frequency. The

inspectors determined that this finding does not have a crosscutting aspect associated

with it since the performance deficiency is not indicative of current licensee performance.

Enforcement. 10 CFR Part 50, Appendix B, Criterion III, Design Control, requires, in

part, that measures be established for the selection and review for suitability of

application of materials, parts, equipment, and processes that are essential to the

safety-related functions of structures, systems, and components. Contrary to the above,

prior to April 2, 2008, the licensee failed to ensure that the containment air coolers would

be able to perform their safety-related function in all accident scenarios due to a design

deficiency associated with the overload contacts in the containment air cooler control

circuitry. Because this finding is of very low safety significance and has been entered

into the corrective action program as CAR 200702264, this violation is being treated as

an NCV consistent with Section VI.A of the NRC Enforcement Policy:

NCV 05000483/2008003-01, Failure to Ensure the Suitability of the Design of the

Containment Air Cooler Control Circuitry.

.2 Introduction. The inspectors identified a Green NCV of Technical Specification 3.5.2,

"Emergency Core Cooling Systems," after an inadequate surveillance procedure

resulted in the licensee failing to maintain the emergency core cooling system (ECCS)

full of water as required per Technical Specification 3.5.2.

Description. On May 21, 2008, Callaway Plant engineering discovered that a section of

the cold leg recirculation piping, specifically the discharge of the residual heat removal

pumps to the safety injection pumps, contained 6.6 cubic feet of air. This exceeded the

allowable void fraction of 2.1 cubic feet required for operability. Callaway monthly

surveillance Procedure OSP-SA-00003, "Emergency Core Cooling Flow Path

Verification and Venting," had a purpose to: "Verify the ECCS is full of water," in

accordance with Technical Specification Surveillance Requirement 3.5.2.3. This

monthly surveillance was reviewed as part of significant condition adverse to quality

(SCAQ) CAR 200501092 corrective actions. Callaway engineering had determined that

residual heat removal pump discharge vent Valve EJV0193 to the safety injection

system was the high point vent for these lines and was thus sufficient to vent

Line EM-023-HCB - 6" to the safety injection pumps. However, this vent valve was not

adequate due to the pipe sloping issues and normally closed Valves EMHIS8807A/B.

Venting through Valve EMV0179 was necessary to completely fill, vent, and test the line.

The monthly verification and vent procedure was inadequate to identify and remove air

- 16 - Enclosure 2

introduced by relief valve maintenance on May 7, 2007, and thus ensure the ECCS was

full of water. See Violation (VIO)05000483/2008003-05 in Section 4OA2.

Analysis. Failure to adequately verify ECCS piping was full of water as required by

Technical Specification 3.5.2 is a performance deficiency. This finding affected the

mitigating system cornerstone procedure quality attribute. This finding is more than

minor because it was similar to Example 3e of NRC Inspection Manual Chapter 0612,

Appendix E, "Examples of Minor Issues," and met the Not Minor If, criteria because the

failure to meet the licensees administrative requirement for allowable void fraction

impacted the ability of the Train A safety injection system to function upon initiation of

high-pressure recirculation. Using Manual Chapter 0609.04, Phase 1 - Initial Screening

and Characterization of Findings, the inspectors determined that this finding should be

evaluated using the Phase 2 process described in Manual Chapter 0609, Appendix A,

Determining the Significance of Reactor Inspection Findings for At-Power Situations.

As described in Section III of Appendix A, given that the presolved table did not contain

a suitable target or surrogate for this finding, the senior reactor analyst used the

risk-informed notebook to evaluate the significance of this finding. Table 2 provides the

definitions for acronyms and initialisms used in the risk-informed notebook and

discussed in this inspection report.

TABLE 2

Acronyms and Initialisms used in Phase 2 Notebook

Initialism Initiating Event or Mitigating Function

TPCS Transient with Loss of the Power Conversion System

SLOCA Small-Break Loss of Coolant Accident

MLOCA Medium-Break Loss of Coolant Accident

LLOCA Large-Break Loss of Coolant Accident

LOOP Loss of Offsite Power

MSLB Main Steam Line Break

LBDC Loss of Vital Direct-Current Bus

AFW Auxiliary Feedwater

PCS Power Conversion System (Steam and Feed)

HPR High Pressure Recirculation

DEPR Depressurization of the Reactor Coolant System

EAC Emergency Power (Alternating Current)

TDAFW Turbine-Driven Auxiliary Feedwater Pump Train

SEAL Reactor Coolant Pump Seal Integrity

STIN Operators Stop High-Pressure Injection

MDAFW Motor-Driven Auxiliary Feedwater Pump Train

The analyst performed a Phase 2 estimation in accordance with Inspection Manual

Chapter 0609, Appendix A, Attachment 2, Site Specific Risk-Informed Inspection

Notebook Usage Rules. Given that the performance deficiency was known to have

existed for 378 days (May 7, 2007, until May 21, 2008) the analyst used 1 year as the

exposure period. In accordance with Table 2 of the risk-informed notebook, the analyst

evaluated all worksheets except LLOCA. All worksheets were evaluated using the

nominal 1-year initiating event frequency. Because this finding only affected system

functionality during recirculation, nominal mitigation credit was given for all functions with

the exception of HPR. For HPR, the analyst made the bounding assumption that either

- 17 - Enclosure 2

both centrifugal charging pumps or both safety injection pumps would be affected. This

assumption was supported by licensee evaluation. The analyst solved each applicable

worksheet and the dominant sequences are documented in Table 1.

TABLE 1

Phase 2 Dominant Sequences

Initiating Event Sequence Mitigating Functions Results

Number

Transients 1 AFW-PCS-HPR 9

TPCS 1 AFW-HPR 8

SLOCA 2 DEPR-HPR 8

MLOCA 2 DEPR-HPR 9

1 AFW-HPR 9

LOOP 5 EAC-TDAFW-HPR 9

9 EAC-SEAL-HPR 9

MSLB 8 STIN-HPR 8

LBDC 8 TDAFW-MDAFW-HPR 8

Using Inspection Manual Chapter 0609, Appendix A, Attachment 1, Table 5, Counting

Rule Worksheet, the analyst determined that the risk contribution of this finding from

internal initiating events was of very low risk significance. In accordance with

Appendix A, Attachment 1, Steps 2.2.5 and 2.2.6, the analyst and determined that the

risk contribution of this finding from external initiating events or the contribution from

large-early release frequency were very low. Therefore, this finding was of very low risk

significance (Green). This finding has a crosscutting aspect in the area of human

performance associated with the decision making component because the licensee

failed to use conservative assumptions in decision making and did not adopt a

requirement to demonstrate that the single vent Valve EJV0193 was sufficient to vent

the Line EM-023-HCB - 6" rather than assuming that installed Valve EMV0179 was not

necessary to completely fill, vent, and test the line H.1(b).

Enforcement. Technical Specification 3.5.2 "Emergency Core Cooling Systems,"

Surveillance Requirement 3.5.2.3, required that the licensee verify that ECCS piping is

full of water every 31 days. Contrary to the above, from June 2007 through April 2008,

AmerenUE surveillance Procedure OSP-SA-00003, "Emergency Core Cooling Flow

Path Verification and Venting," was inadequate to meet Technical Specification

Surveillance Requirement 3.5.2.3. Because this finding is of very low safety significance

and was entered into the licensee's corrective action program as CAR 200804000, this

violation is being treated as an NCV, consistent with Section VI.A.1 of the NRC

Enforcement Policy: NCV 05000483/2008003-02, Inadequate Surveillance Procedure

Resulted in an Inoperable ECCS.

- 18 - Enclosure 2

1R18 Plant Modifications (71111.18)

a. Inspection Scope

The inspectors reviewed the design adequacy of the listed modifications. This included

verifying that the modification preparation did not impair the following: (a) in-plant

emergency/abnormal operating procedure actions, (b) key safety functions, and

(c) operator response to loss of key safety functions.

The inspectors verified that postmodification testing maintained the plant in a safe

configuration during testing and that the postmodification testing established operability

by: (a) verifying that unintended system interactions did not occur; (b) verifying that

performance characteristics, which could have been affected by the modification, met

the design bases; (c) validating the appropriateness of modification design assumptions;

and (d) demonstrating that the modification test acceptance criteria had been met.

• April 18, 2008, Modification M-08-0013 to separate fast and slow speed overload

contacts for containment air coolers

• June 1, 2008, Temporary Modification TM 08-0003 for the instrument air system

to provide an additional diesel-driven air compressor to improve system reliability

while the system was in degraded reliability

Documents reviewed are listed in the attachment.

These activities constituted two samples as defined by Inspection Procedure 71111.18.

b. Findings

No findings of significance were identified

1R19 Postmaintenance Testing (71111.19)

a. Inspection Scope

The inspectors reviewed the following postmaintenance activities to verify that

procedures and test activities were adequate to ensure system operability and functional

capability:

• April 10, 2008, Job 08002765.900, Source Range N32 postmaintenance test

• April 17, 2008, Postmaintenance test containment Cooler D,

Modification 0800267/950(951)(952)

• May 7, 2008, Job 06524419.940, Emergency Diesel Generator B

• May 28, 2008, Job 08003910, Postmaintence test of Emergency Diesel

Generator A following repair of jacket water leaks

• May 30, 2008, Job 08001080, Postmaintenance local leakrate test of

containment personnel hatch door

- 19 - Enclosure 2

These activities were selected based upon the structure, system, and component's

ability to impact risk. The inspectors evaluated these activities to verify (as applicable):

the effect of testing on the plant had been adequately addressed; testing was adequate

for the maintenance performed; acceptance criteria were clear and demonstrated

operational readiness; test instrumentation was appropriate; tests were performed as

written in accordance with properly reviewed and approved procedures; equipment was

returned to its operational status following testing (temporary modifications or jumpers

required for test performance were properly removed after test completion); and test

documentation was properly evaluated. The inspectors evaluated the activities against

Technical Specifications, the FSAR, 10 CFR Part 50 requirements, licensee procedures,

and various NRC generic communications to ensure that the test results adequately

ensured that the equipment met the licensing basis and design requirements. In

addition, the inspectors reviewed corrective action documents associated with

postmaintenance tests to determine whether the licensee was identifying problems and

entering them in the corrective action program and that the problems were being

corrected commensurate with their importance to safety. Documents reviewed are listed

in the attachment.

This inspection constitutes five samples as defined in Inspection Procedure 71111.19.

b. Findings

Introduction. A self-revealing Green NCV of 10 CFR Part 50, Appendix B, Criterion XVI,

"Corrective Action," was identified after the licensee failed to promptly correct leakage

from diesel generator jacket water o-rings.

Description. On February 20, 2008, during performance of Procedure OSP-NE-0001B,

Standby Diesel Generator B Periodic Tests, Callaway operations personnel identified

that the Emergency Diesel Generator B had an approximately 80 drop-per-minute jacket

water leak. Analysis by the licensee determined the cause of the leakage to be from

premature failure of Nitrile type o-rings in the jacket water supply and return headers.

Operational history at Callaway revealed o-ring failures prior to reaching 3 years of

service life. The o-rings responsible for the February 20, 2008, leakage had been in

service since Refueling Outage 14 in October 2005. Following restoration of Emergency

Diesel Generator B, the licensee re-evaluated the preventative maintenance frequency

for jacket water o-ring replacement. Based on a review of prior o-ring failures, the

replacement schedule for diesel generator jacket water o-rings was reduced from once

every 3 years to once every refueling cycle.

On May 28, 2008, during performance of Procedure OSP-NE-0001A, Standby Diesel

Generator A Periodic Tests, Callaway operations personnel identified that Emergency

Diesel Generator A had a 200 drop-per-minute jacket water leak. Based on the quantity

of the leakage, operations personnel declared Emergency Diesel Generator A

inoperable. Similar to the condition observed on Emergency Diesel Generator B on

February 20, 2008, the source of the leakage was from Nitrile type o-rings within the

jacket water system. While the licensee replaced the o-rings responsible for jacket

water leakage following the February 20, 2008, surveillance, several Nitrile type o-rings

installed during Refueling Outage 14 in October 2005 remained in service in both

Emergency Diesel Generators Trains A and B including those that failed during the

May 28, 2008, surveillance.

- 20 - Enclosure 2

Subsequent analysis by the licensee determined that the required mission time of the

Emergency Diesel Generator A was preserved since adequate inventory in the jacket

water expansion tank existed such that the leakage observed on May 28, 2008, would

not have impacted the net positive suction head analysis for the jacket water cooling

pump.

Analysis. The performance deficiency associated with this finding involved the

licensees failure to implement adequate corrective actions for an adverse condition.

Specifically, the licensee failed to correct degraded Nitrile type o-rings in Emergency

Diesel Generator A after previously identifying the adverse condition on Emergency

Diesel Generator B. This finding was greater than minor because, if left uncorrected,

degraded diesel generator jacket water o-rings could become a more significant safety

concern. This finding affected the mitigating systems cornerstone. Using Manual

Chapter 0609.04, Phase 1 - Initial Screening and Characterization of Findings, this

finding was determined be of very low safety significance because it was a design

deficiency confirmed not to result in loss of operability. This finding had a crosscutting

aspect in the area of human performance associated with the work control component

because the licensee failed to plan work activities to support long-term equipment

reliability by addressing known degraded conditions in a more reactive than preventative

manner H.3(b).

Enforcement. 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, requires,

in part, that measures be established to assure conditions adverse to quality are

promptly identified and corrected. Contrary to the above, the licensee failed to

implement adequate corrective actions for the identified adverse condition that Nitrile

type o-rings would prematurely fail prior to the completion of the regularly scheduled

3-year replacement interval. Because this violation is of very low safety significance and

has been entered into the licensee's corrective action program as CAR 200804164, this

violation is being treated as an NCV, consistent with Section VI.A.1 of the NRC

Enforcement Policy: NCV 05000483/2008003-03, Failure to Correct a Condition

Adverse to Quality for Diesel Generator Jacket Water O-Rings.

1R22 Surveillance Testing (71111.22)

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:

• April 2, 2008, Job 07513515.500, Routine surveillance auxiliary building Train A

negative pressure test

• April 4, 2008, Job 08501501, Routine surveillance Slave Relay K645 test of

essential service water component lineup

• April 18, 2008, Job 08501499, Routine surveillance Slave Relay K615 test

• April 29, 2008, Job 08501254.500, Residual heat removal Pump A inservice test

- 21 - Enclosure 2

• May 5, 2008, Jobs 08502271 and 08503327, Routine surveillance containment

base strong motion accelerometer seismic monitor calibration

• May 14, 2008, Job 07505653, Residual heat removal Train B valve inservice test

• June 11, 2008, Job 08504820, Routine surveillance diesel generator Train B

1-hour run

• June 17, 2008, Job 08503115, Safety injection system Train A valve inservice

test

• June 18, 2008, Job 08505155, Routine surveillance ECCS flow path verification

and venting

• June 23, 2008, Job 08506247, Reactor coolant system leakage surveillance,

reactor coolant system inventory balance, plant status

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 FSAR, 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; 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.

Documents reviewed are listed in the attachment.

The inspectors completed six routine, three inservice test, and one reactor coolant

system leakage samples.

b. Findings

Introduction. A self-revealing Green NCV of Technical Specification 5.4.1.a,

Procedures, was identified after Callaway control room operators improperly entered

the wrong Technical Specification action statement due to the failure to maintain the

Technical Specification Bases current.

- 22 - Enclosure 2

Description. On June 17, 2008, during surveillance testing, Valve EMHV8823 failed to

indicate fully closed. Since EMHV8823 is an isolation valve for containment

Penetration 49, the licensee entered Technical Specification 3.6.3, Containment

Isolation Valves," Condition C, with an action to restore the valve to an operable status

or isolate the penetration within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. The control room staff believed the

appropriate action statement was entered since Condition C is described in the

Technical Specification Bases as applicable to flow paths that meet the requirements of

a closed system per the Callaway FSAR. Chapter 6.2.6.3 of the Callaway FSAR

described Containment Penetration 49 as a closed engineered safety feature

containment penetration.

Approximately 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> after Valve EMHV8823 had been declared inoperable, Callaway

licensing personnel contacted the control room and informed them of an approved

Technical Specification Bases change that did not allow the classification of containment

Penetration 49 as a closed system. Procedure APA-ZZ-00108, Primary Licensing

Document; Change/Revision Process," required that the change be implemented within

45 days following approval. The Technical Specification Bases change was effective

May 1, 2008, but had not been issued to the control room. The change resulted in

Condition C of Technical Specification 3.6.3 applying specifically to penetrations for

which a single containment isolation valve is credited per flow path. Since containment

Penetration 49 relies on multiple valves for flow path isolation, the licensee determined

that Condition C of Technical Specification 3.6.3 was not applicable for Penetration 49,

and the wrong Technical Specification action statement had been entered following the

failed surveillance on Valve EMHV8823. The licensee determined that the more

restrictive Technical Specification 3.6.3, Condition A, should have been entered with an

action to isolate the affected penetration within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

The licensee performed a containment entry following discovery of entry into Technical Specification 3.6.3, Condition A, and found that Valve EMHV8823 had failed its

surveillance due to out-of-adjustment position indicator limit switches. The valve was

verified closed with power removed allowing exit from Technical Specification 3.6.3,

Condition A.

Analysis. The performance deficiency associated with this finding involved the

licensees failure to ensure the Technical Specification Bases were maintained current

and available to the Callaway control room staff. This finding was greater than minor

because, if left uncorrected, the failure to maintain the Technical Specification Bases

current could become a more significant safety concern. This finding was determined to

affect the barrier integrity cornerstone. Using Manual Chapter 0609.04, Phase 1 - Initial

Screening and Characterization of Findings," this finding is determined to be of very low

safety significance since this finding did not represent an actual open pathway in the

physical integrity of reactor containment and did not involve an actual reduction in

function of hydrogen ignitors in the reactor containment. This finding had a crosscutting

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

component because the licensee failed to communicate, in a timely manner, decisions to

personnel who have a need to know the information in order to perform work safely

H.1(c).

Enforcement. Technical Specification 5.4.1.a, Procedures, required that written

procedures be established and implemented covering activities specified in Appendix A,

Typical Procedures for Pressurized Water Reactors, of Regulatory Guide 1.33, Quality

- 23 - Enclosure 2

Assurance Program Requirements (Operation), February 1978. Regulatory Guide 1.33,

Appendix A, Section 1, required administrative procedures for procedure review and

approval. Procedure APA-ZZ-00108 provides a process for implementing Technical

Specification Bases change notices. Contrary to the above, on May 1, 2008,

Procedure APA-ZZ-00108 was not adequate to ensure changes to the Technical

Specification Bases were implemented in a timely manner. Because of the very low

safety significance and AmerenUEs action to place this issue in their corrective action

program as CAR 200805283, this violation is being treated as an NCV in accordance

with Section VI.A.1 of the Enforcement Policy: NCV 05000483/2008003-04, Failure to

Maintain an Adequate Technical Specification Bases Change Process.

2. RADIATION SAFETY

Cornerstone: Occupational Radiation Safety

2OS1 Access Control to Radiologically Significant Areas (71121.01)

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

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

airborne radioactivity areas

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

locations

• Physical and programmatic controls for highly activated or contaminated

materials (non-fuel) stored within spent fuel and other storage pools

• Self-assessments, audits, LERs, and special reports related to the access control

program since the last inspection

• 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 accessible high dose rate - high radiation

areas and very high radiation areas

- 24 - Enclosure 2

Documents reviewed are listed in the attachment.

The inspectors completed 8 of the required 21 samples.

b. Findings

No findings of significance were identified.

2OS2 ALARA Planning and Controls (71121.02)

a. Inspection Scope

The inspectors assessed licensee performance with respect to maintaining individual

and collective radiation exposures as low as is reasonably achievable (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

inspectors interviewed licensee personnel and reviewed:

• Current 3-year rolling average collective exposure

• Site-specific trends in collective exposures, plant historical data, and source-term

measurements

• Site-specific ALARA procedures

• Work activities of highest exposure significance during the inspection

• Integration of ALARA requirements into work procedure and radiation work

permit documents

• Post-job (work activity) reviews

• Workers use of the low dose waiting areas

• First-line job supervisors contribution to ensuring work activities are conducted in

a dose efficient manner

• Records detailing the historical trends and current status of tracked plant source

terms and contingency plans for expected changes in the source term due to

changes in plant fuel performance issues or changes in plant primary chemistry

• Source-term control strategy or justifications for not pursuing such exposure

reduction initiatives

• Specific sources identified by the licensee for exposure reduction actions,

priorities established for these actions, and results achieved since the last

refueling cycle

• Radiation worker and radiation protection technician performance during work

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

- 25 - Enclosure 2

• Declared pregnant workers during the current assessment period, monitoring

controls, and the exposure results

• 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

post-job reviews and post-outage ALARA report critiques

• Corrective action documents related to the ALARA program and follow-up

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

• Effectiveness of self-assessment activities with respect to identifying and

addressing repetitive deficiencies or significant individual deficiencies

Documents reviewed are listed in the attachment.

The inspectors completed 9 of the required 15 samples and 8 of the optional samples.

b. Findings

No findings of significance were identified.

4. OTHER ACTIVITIES

4OA1 Performance Indicator Verification (71151)

.1 Data Submission Issue

a. Inspection Scope

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

Quarter 2008 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 Safety System Functional Failures

Cornerstone: Mitigating Systems

a. Inspection Scope

The inspectors sampled licensee submittals for the safety system functional failures

performance indicator for the period March 2007 until March 2008. To determine the

accuracy of the performance indicator data reported during this period, performance

indicator definitions and guidance contained in the Nuclear Energy Institute (NEI)

- 26 - Enclosure 2

Document 99-02, Revision 5, Regulatory Assessment Performance Indicator

Guideline, and NUREG-1022, Event Reporting Guidelines 10 CFR 50.72 and 50.73,"

definitions and guidance were used. The inspectors reviewed the licensees operator

narrative logs, operability assessments, maintenance rule records, maintenance work

orders, issue reports, event reports and NRC integrated inspection reports for the period

of 2nd Quarter 2007, through 1st Quarter 2008 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

are listed in the attachment.

This inspection constitutes one safety system functional failures sample as defined by

Inspection Procedure 71151.

b. Findings

No findings of significance were identified.

.3 Mitigating Systems Performance Index - High Pressure Injection Systems

Cornerstone: Mitigating Systems

a. Inspection Scope

The inspectors sampled licensee submittals for the mitigating systems performance

index - high pressure injection systems performance indicator for the period from

March 2007 until March 2008. To determine the accuracy of the performance indicator

data reported during this period, performance indicator definitions and guidance

contained in the NEI Document 99-02, 5, Regulatory Assessment Performance

Indicator Guideline, Revision 5, were used. The inspectors reviewed the licensees

operator narrative logs, issue reports, mitigating systems performance index derivation

reports, event reports, and NRC integrated inspection reports for the period of

2nd Quarter 2007 through 1st Quarter 2008 to validate the accuracy of the submittals.

The inspectors reviewed the mitigating systems performance index component risk

coefficient to determine if it had changed by more than 25 percent in value since the

previous inspection, and if so, that the change was in accordance with applicable NEI

guidance. 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

are listed in the attachment.

This inspection constitutes one mitigating systems performance index high pressure

injection systems sample as defined by Inspection Procedure 71151.

b. Findings

No findings of significance were identified.

- 27 - Enclosure 2

.4 Occupational Exposure Control Effectiveness

Cornerstone: Occupational Radiation Safety

a. Inspection Scope

The inspectors reviewed licensee documents from October 1, 2007, through March 31,

2008. 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 inspectors interviewed licensee personnel

that were accountable for collecting and evaluating the performance indicator data. In

addition, the inspectors 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.

The inspectors completed the required sample (1) in this cornerstone.

b. Findings

No findings of significance were identified.

.5 Radiological Effluent Technical Specification/Offsite Dose Calculation Manual

Radiological Effluent Occurrences

Cornerstone: Public Radiation Safety

a. Inspection Scope

The inspectors reviewed licensee documents from October 1, 2007, through March 31,

2008. 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 inspectors 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.

The inspectors completed the required sample (1) in this cornerstone.

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

- 28 - Enclosure 2

.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

to verify that they were being entered into the licensees corrective action program at an

appropriate threshold, that adequate attention was being given to timely corrective

actions, and that adverse trends were identified and addressed. The 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 occurrence 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.

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

any additional inspection samples.

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 inspection 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.

b. Findings

No findings of significance were identified.

.3 Selected Issue Follow-up Inspection

a. Inspection Scope

The inspectors selected the below listed issues for a more in-depth review. The

inspectors considered the following during the review of AmerenUE'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.

- 29 - Enclosure 2

• Voiding discovered in the common residual heat removal discharge piping for

high pressure recirculation.

• FSAR changes/updates

Documents reviewed are listed in the attachment.

This inspection constituted two in-depth problem identification and resolution samples.

b. Findings

Introduction. The inspectors identified a Green violation of 10 CFR Part 50, Appendix B,

Criterion XVI, "Corrective Action," because the licensee failed to take corrective actions

to preclude repetition of void formations in the ECCS, a significant condition adverse to

quality (SCAQ). Contributors to the violation included: (1) the failure of corrective

actions from inspection report findings NCV 05000483/2005002-01,

05000483/2006012-04 and CAR 200501092 to ensure adequate fill and vent of

systems following maintenance to replace safety injection system relief valves, and

(2) inadequate extent of condition reviews in responding to internal and external

operating experience associated with pipe sloping issues in the safety injection system.

Description. On May 21, 2008, the Callaway Plant staff initiated CAR 200804000, a

SCAQ corrective action document, indicating that some piping in Train A safety injection

system suction lines had incorrect sloping and were susceptible to voiding due to high

points. Callaway Plant engineering performed ultrasonic inspection of the safety

injection system common suction piping Line EM023-HCB - 6" and discovered a

6.6 cubic foot voided area. This exceeded the allowable void fraction of 2.1 cubic feet

required for operability. This voided piping, determined to have existed for over a year,

was caused by relief valve maintenance on Valve EM8858A (May 7, 2007). The

maintenance restoration failed to perform an adequate fill and vent to ensure the suction

pipe was full of water.

In 2005 and 2006 the NRC issued NCVs regarding ineffective corrective actions related

to safety injection system voids in discharge piping (05000483/2005002-01 dated May 6,

2005, and 05000483/2006012-04 dated December 26, 2006). These were each 10 CFR

Part 50, Appendix B, Criterion XVI, NCVs, each for SCAQ. The Callaway Plant staff

issued CAR 200501092 as a SCAQ corrective action document. The CAR determined

that the causes of the voids (2004, 2005, and 2006) were related to incorrect pipe

sloping (allowing high points where voids could not be swept away by normal online

pump surveillances) and inadequate postmaintenance fill and vent operations (following

discharge piping relief Valve EM8853A replacement) to ensure the piping was full of

water.

Inadequate Operating Experience and Extent of Condition Corrections: The

inspectors identified several related examples where the licensee had performed either

inadequate operating experience evaluations, inadequate extent of condition reviews, or

inadequate procedure corrections.

Callaway CAR 200501092 referenced industry operating experience at Beaver Valley

Unit 2 in 2002: "The void was located in the piping used following a loss of coolant

- 30 - Enclosure 2

accident after the transfer to containment sump recirculation. The piping containing the

void led to a common suction header for both trains of high head pumps." This was the

same location as the voiding discovered at Callaway Plant on May 21, 2008.

NRC Information Notice 2006-21, "Operating Experience Regarding Entrainment of Air

into Emergency Core Cooling and Containment Spray Systems," dated September 21,

2006, discussed mechanisms that could result in air entrainment on the suction sides of

emergency core cooling pumps. The notice emphasized the importance of ensuring that

entrained air will not enter suction supply lines and impair the ability of the ECCS and

containment spray pumps to perform their safety function.

The licensee's evaluation of NRC Information Notice 2006-21 was documented in

CAR 200608956. It stated that the information notice was applicable to Callaway and

that past review of these operating experiences and Callaway procedures and practices

were adequate. The CAR was closed December 5, 2006.

Callaway CAR 200501092 had Action 7 assigned to address the previous NRC

violations discussed above. The action required that system specific fill and vent

restoration guidance be developed to address maintenance on ECCS safety-related

systems. Initially, operating department Standing Order 05-002 dated June 8, 2005,

stated that the CAR 200501092 common cause analysis supported the need for

formalized restoration instructions. Until the system specific restoration instructions

were developed, the standing order required reactor operators to perform reviews to

ensure dynamic filling and venting occurred to reduce the susceptibility of voiding. Also

nuclear engineering department staff were to provide concurrence on such restoration

plans. Night Order ODP-ZZ-00310, "System Fill and Vent," issued February 13, 2006,

reiterated that reactor operator reviews and engineering concurrence were required

when these risk-significant systems were drained. However, on May 7, 2007,

Procedure OTN-EM-00001, "Safety Injection System," (developed to address filling and

venting evaluations) had Line EM-023-HCB - 6" isolated by Valve EMHIS8807B being

closed. The procedure did not include use of the available installed vent Valve EM179

for this line.

Callaway monthly Surveillance OSP-SA-00003, "Emergency Core Cooling Flow Path

Verification and Venting," had a purpose to: "Verify the ECCS is full of water in

accordance with Technical Specification Surveillance Requirement 3.5.2.3." This

monthly surveillance was reviewed as part of CAR 200501092 corrective actions.

Callaway engineering had determined that residual heat removal pump discharge vent

Valve EJV0193 to the safety injection suction line was the high point vent for these lines

and was thus sufficient to vent supply Line EM-023-HCB - 6" to the safety injection

pumps. However, this vent valve was not adequate due to the pipe sloping issues and

normally closed Valves EMHIS8807A/B. The monthly verification and vent procedure

was inadequate to remove the air entrained by the May 7, 2007, relief valve

maintenance. See Section 1R15, NCV 05000483/2008003-02.

Callaway CARs 200800226 and 200800246, initiated in January 2008, discussed

operating experience at Wolf Creek Nuclear Operating Corporation describing gas

voiding in the residual heat removal piggyback Line EM-022-HCB - 6" to the suction of

centrifugal charging pumps and safety injection pumps. The CARs stated that Callaway

had taken a proactive approach and had immediately performed ultrasonic testing to

demonstrate that the associated piping was water solid. However, the adjacent

- 31 - Enclosure 2

connecting Line EM-023-HCB - 6" had not been vented nor had ultrasonic testing

occurred since the May 7, 2007, relief Valve EM8858A maintenance.

NRC Generic Letter 2008-01, "Managing Gas Accumulation in Emergency Core Cooling,

Decay Heat Removal, and Containment Spray Systems," was issued January 11, 2008.

The Callaway Plant staff initiated CAR 200800298 to respond to the generic letter. The

generic letter identified that a licensing basis concern existed for some plants, such as

Callaway, that Technical Specifications require verifying that ECCS discharge piping is

full of water but may not include verification of the suction piping despite the realistic

concern that gas accumulation in suction piping may be more serious than gas

accumulation in discharge piping. The void found in Line EM-023-HCB - 6" was the

discharge of the residual heat removal pumps providing suction to the Train A safety

injection pump. The Callaway monthly Surveillance OSP-SA-00003, "Emergency Core

Cooling Flow Path Verification and Venting," did not test for or vent the discharge line

from residual heat removal to safety injection pump suction piping.

Analysis. The inspectors determined that the failure to restore compliance within a

reasonable time by establishing measures to prevent void formation in ECCS suction

piping for the Train A safety injection system was a performance deficiency. This finding

is more than minor because it was similar to Example 3e of NRC Inspection Manual

Chapter 0612, Appendix E, "Examples of Minor Issues," and met the Not Minor If,

criteria because the failure to meet the licensees administrative requirement for

allowable void fraction impacted the ability of the Train A safety injection system to

function upon initiation of high-pressure recirculation. Using Manual Chapter 0609.04,

Phase 1 - Initial Screening and Characterization of Findings, the inspectors determined

that this finding should be evaluated using the Phase 2 process described in Manual

Chapter 0609, Appendix A, Determining the Significance of Reactor Inspection Findings

for At-Power Situations.

The senior reactor analyst determined that the risk of this finding was bounded by that

analyzed for NCV 05000423/2008003-02 (See Section 1R15.b.2). Therefore, this

finding was of very low risk significance (Green).

This finding has a crosscutting aspect in the area of problem identification and resolution

associated with the corrective action component because AmerenUE failed to thoroughly

evaluate voiding problems such that the resolutions addressed causes and extent of

condition, as necessary. This also includes, for significant problems, conducting

effectiveness reviews of corrective actions to ensure that the problems are resolved

P.1(c).

Enforcement. 10 CFR Part 50, Appendix B, Criterion XVI, "Corrective Action," requires

the licensee to, in the case of SCAQ, establish measures to assure that the cause of the

condition is determined and corrective action is taken to preclude repetition. Contrary to

the above, from December 26, 2006, to May 21, 2008, the licensee did not implement

corrective action to preclude repetition of void formation in the safety injection piping

which the licensee categorized as an SCAQ. Specifically, void formation recurred after

performing maintenance on relief valve. Valve EM8858A, on May 7, 2007. Previously

discovered voiding of the safety injection system was last documented as an SCAQ in

NCV 05000483/2006012-04 dated December 26, 2006. For each instance of the

previously discovered voids, the causes were determined to be related to inadequate fill

and vent of the system piping following relief valve replacements and design deficiencies

- 32 - Enclosure 2

associated with inadequate sloping of the piping. It was a reasonable assumption that

maintenance that drained either the suction or discharge piping could create significant

void areas.

Although this violation is of very low safety significance, the violation is being

cited in a Notice of Violation consistent with Section VI.A.1 of the NRC Enforcement

Policy because the licensee did not restore compliance within a reasonable

time after a previous violation NCV 05000483/2006012-04 was identified:

VIO 05000483/2008003-05, Failure to Prevent Recurrence of Voids in ECCS Cold Leg

Recirculation Piping. This finding has been entered into the licensee's corrective action

program as a SCAQ in CAR 200804000.

.4 Semiannual Trend Review

The inspectors assessed trends that might indicate the existence of a more significant

safety issue. These issues included trends that might not rise to the level of an

inspection finding.

NRC-Identified Trends

The NRC identified emergency diesel generator material condition and design control

issues degrading diesel reliability:

• CAR 200801270: 80 Drops per minute jacket water leak identified on Diesel

Generator B

• CAR 200801644: Additional sacrificial anode found in Emergency Diesel

Generator A intercooler heat exchanger

• CAR 200802019: Emergency Diesel Generator B declared inoperable due to

fuel oil leaks

• CAR 200802177: Cracked fuel oil return line fitting identified on Emergency

Diesel Generator A

• CAR 200804164: Emergency Diesel Generator A declared inoperable due to a

200 drops per minute jacket water leak

Licensee-Identified Trends

The licensee identified a continued trend in plant status control and configuration control

with a key causal factor being procedure adherence.

• CAR 200706832: This trend CAR from Third Quarter 2007 identified the cause

of plant status control issues to be a "Failure to follow written instructions."

• CAR 200801457: A gauge was installed on an incorrect component during Test

Procedure OSP-EN-P001A.

• CAR 200800580: A trend of critical steps not being included in work packages

was identified.

- 33 - Enclosure 2

• CAR 200802603: Component cooling water pump autostarted due to an

interlock with the centrifugal charging pumps. The operator failed to wait the

procedure prerequisite 30 minutes prior to securing the component cooling water

pump.

• CAR 200802818: Source range Channel N31 was not restored to "block" as

required by procedure in Mode 1.

• CAR 200800328: Not following procedures resulted in gaseous Radiation

Monitor RM-11 trip setpoint not being capable of isolating the waste gas decay

tank release.

• CAR 200803351: Steam generator blowdown tripped due to an incorrect

demineralizer valve lineup.

• CAR 200804483: Train B motor-driven auxiliary feedwater pump made

inoperable when its room cooler was taken to "stop" vice "auto." This was

performed outside the out of service restoration process.

This inspection constituted one semiannual trend review sample.

4OA3 Event Follow-up (71153)

(Closed) LER 05000483/2008-001-00, Containment Cooler Inoperability

On March 26, 2008, Containment Air Cooler A fan shut down when shifted from fast to

slow speed. The licensee determined that operation of containment air coolers in fast

speed, during a period of higher than normal containment pressure, would challenge the

fast speed thermal overload setpoint. Additionally, since the overload contacts are wired

in series, containment air coolers were determined to experience a complete loss of

control power following a trip from fast speed. The licensee analyzed the potential

impact of the containment cooler design vulnerability against design basis accident

scenarios. The licensee determined that a hot zero power main steam line break results

in a delayed safety injection signal allowing the fan motor overloads to trip prior to being

shed from the load sequencer. In this scenario, utilizing actual plant conditions, the peak

containment pressure would not exceed the 48.1 psig limit described in the FSAR. To

address the design deficiency associated with the containment air cooler control

circuitry, the licensee completed a modification in April 2008 to reconfigure the circuit

such that tripping of the fast speed overloads would not impact the safety-related slow

speed function of the containment air coolers. This finding is of very low safety

significance because the containment coolers are structures, systems, and components

that are not significant contributors to the large early release frequency. Licensee

corrective actions were recorded in CAR 200802264. The inspectors reviewed the LER

and identified a Green NCV of 10 CFR Part 50, Appendix B, Criterion III, Design

Control, for the licensees failure to adequately review the suitability of the design of the

containment air cooler control circuitry (Section 1R15). This LER is closed.

This inspection constituted one sample of follow-up of events.

- 34 - Enclosure 2

4OA5 Other Activities

.1 Quarterly Resident Inspector Observations of Security Personnel and Activities

a. During the inspection period, the inspectors performed the following observations of

security force personnel and activities to ensure that the activities were consistent with

licensees security procedures and regulatory requirements relating to nuclear plant

security. These observations took place during both normal and off-normal plant

working hours.

These quarterly resident inspector observation of security force personnel and activities

did not constitute any additional inspection samples. Rather, they were considered an

integral part of the inspectors normal plant status review and inspection activities.

b. Findings

No findings of significance were identified.

.2 (Closed) NRC Temporary Instruction 2515/166: Pressurized Water Reactor

Containment Sump Blockage

a. Inspection Scope

From March 17-19, 2008, the inspectors reviewed the licensees implementation of plant

modifications and design modification packages associated with their response to

Generic Letter 2004-02, Potential Impact of Debris Blockage on Emergency

Recirculation During Design Basis Accidents at Pressurized Water Reactors. The

inspectors reviewed various aspects of the on-going procedural changes. Those

changes that have been completed were verified to be properly documented in

accordance with the requirements of 10 CFR 50.59. At the completion of this inspection,

the licensee had completed the installation stage of the new sump strainers; many of the

procedural changes associated with the modifications had not been completed.

The inspectors compared and evaluated the recirculation sump modifications to the

original design basis using Temporary Instruction 2515/166 and referred to Regulatory

Guide 1.82, Revision 0, Water Sources for Long-Term Recirculation Cooling Following

a Loss-of-Coolant Accident.

Status of the implementation of the plant modifications and procedure changes

committed to by the licensee in their Generic Letter 2004-02 response is:

1. Containment walkdown to provide current assessment of Callaway's containment

coatings and latent debris.

The licensee completed a containment walkdown and latent debris assessment

during Refueling Outage 14. The resident inspectors completed a walkdown of

the containment prior to reactor startup following the outage. The licensee

report, Containment Building Latent Debris Assessment Refuel 14 Fall 2005,

was reviewed by the inspectors.

- 35 - Enclosure 2

2. The following corrective action activities will be completed:

a. Replacement sump strainer structural analysis.

The strainers were not built in accordance with the design. As a result,

calculations needed to be revised due to the deviations of the as built

condition from design and errors in temperature correction values used in

the initial calculations. Completion date: June 30, 2008

b. Downstream effects evaluation

Completion date: June 30, 2008

c. Upstream effects evaluation

Completion date: June 30, 2008

d. Resolution of debris generation calculation unverified assumption of 5D

ZOI for qualified coatings (via coatings testing)

Completion date: June 30, 2008

e. Replacement sump screen head loss testing

Completion date: June 30, 2008

3. Provide an update of the information contained in Section 2(c) regarding analysis

methodology.

Completion date: June 30, 2008

4. The following evaluations and testing will be completed.

a. Industry chemical effects testing

Completion date: June 30, 2008

b. Nuclear Energy Institute 04-07 debris generation calculation

Completion date: June 30, 2008

c. Evaluation of chemical effects impact on sump-strainer head loss

Completion date: June 30, 2008

d. Confirmation that the replacement sump strainer design provides for

available Net Positive Suction Head (NPSH) to be in excess of required

NPSH

Completion date: June 30, 2008

- 36 - Enclosure 2

e. Completion of the final site acceptance review of the Westinghouse team

analysis summary report

Completion date: June 30, 2008

5. Callaway Plant will complete the following items during Refueling Outage15:

a. Replacement of containment recirculation sump strainers

Completed. As noted in the previous Temporary Instruction 166 report,

the resident inspectors had observed the installation of sump strainers

and debris barriers during their containment walkdown; however, the

strainers were not built in accordance with the design. The licensee has

completed their initial determination of operability and was finalizing their

acceptance calculations.

b. Modification of containment debris barriers and interceptors as required

Completed. As noted in the previous Temporary Instruction 166 report,

the resident inspectors had observed the installation of sump strainers

and debris barriers during their containment walkdown.

c. Evaluation and implementation of potential modification to the safety

injection system to address downstream effects

Completion date: June 30, 2008

6. Callaway Plant will complete removal of containment spray system pump cyclone

separators, if required, based on the results of the downstream effects

evaluation.

Completion date: June 30, 2008

7. The following programs and controls will be implemented at Callaway Plant to

control debris sources:

a. Changes to design change process procedures to ensure that necessary

engineering evaluations will be performed for plant design that either

directly or indirectly affects containment, ECCS, or CSS.

Changes are being processed.

b. Changes to containment entry and material control procedure

requirements for control of materials during work activities conducted in

the containment

c. The following procedures were reviewed and completed as of

December 2007:

APA-ZZ-01004, Radiological Work Standards, Revision 9

HDP-ZZ-06100, Reactor Building Access, Revision 7

- 37 - Enclosure 2

MDP-ZZ-S0001, Scaffolding Installation and Evaluation, Revision 22

OSP-EJ-00003, Containment Recirculation Sump Inspection, Revision 6

OSP-SA-00004, Visual Inspection of Containment for Loose Debris,

Revision 19

OTS-ZZ-06032, Addendum 1, Debris and Particulate Control Devices,

Revision 2

d. Changes to programs and procedures that have the potential to add tags

and labels inside containment

Completed: December 2007

The following documents were reviewed:

APA-ZZ-01004, Radiological Work Standards, Revision 9

HDP-ZZ-06100, Reactor Building Access, Revision 7

MDP-ZZ-S0001, Scaffolding Installation and Evaluation, Revision 22

OSP-EJ-00003, Containment Recirculation Sump Inspection, Revision 6

OSP-SA-00004, Visual Inspection of Containment for Loose Debris,

Revision 19

OTS-ZZ-06032, Addendum 1, Debris and Particulate Control Devices,

Revision 2

e. Implementation of a containment coatings assessment program

Licensee reported as complete. The inspectors reviewed SWE07848,

Containment Coating Condition Assessment. A preventative

maintenance item has been scheduled to perform containment coating

assessments with a periodicity of each refueling cycle.

f. Implementation of a containment latent debris assessment program

Licensee reported as complete. The inspectors reviewed report,

Containment Building Latent Debris Assessment Refuel 14 Fall 2005,

and Procedure OSP-SA-00004, Visual Inspection of Containment for

Loose Debris, Revision 019. A preventative maintenance item has been

scheduled for a visual inspection of containment for loose debris with a

periodicity of each refueling cycle.

g. Implementation of changes to the inspection processes for the installed

sump strainers

Licensee reported as complete. Reviewed Procedure OSP-EJ-00003,

Containment Recirculation Sump Inspection, Revision 6

- 38 - Enclosure 2

8. A final response will be submitted to the NRC to provide a final status of actions

requested by Generic Letter 2004-02.

Completion date: June 30, 2008

The Office of Nuclear Reactor Regulation will determine the adequacy of the sump

modifications with respect to Generic Safety Issue 191. This temporary instruction is

closed.

Documents reviewed by the inspectors are listed in the attachment.

b. Findings

No findings of significance were identified.

4OA6 Management Meetings

Exit Meeting Summary

On April 25, 2008, the health physics inspector presented the occupational radiation

safety inspection results to Mr. T. Herrmann and other members of his staff who

acknowledged the findings. The inspector confirmed that proprietary information was

not provided or examined during the inspection.

On June 18, 2008, the Temporary Instruction 2515/166 inspector presented the

inspection results to Mr. S. Maglio and other members of his staff who acknowledged the

findings. The inspector confirmed that proprietary information provided or examined

during the inspection had been returned.

On June 24, 2008, the resident inspectors presented the inspection results to

Mr. C. Naslund, Senior Vice President and Chief Nuclear Officer, and other members of

the licensee staff. The licensee acknowledged the issues presented. The inspectors

understood and acknowledged that proprietary information reviewed would not be

retained following report issuance.

4OA7 Licensee-Identified Violations

The following violations of very low safety significance (Green) were identified by the

licensee and were violations of NRC requirements which meet the criteria of Section VI

of the NRC Enforcement Policy, NUREG-1600, for being dispositioned as NCVs.

• 10 CFR Part 50, Appendix B, Criterion III, Design Control, requires, in part,

that applicable regulatory requirements and the design basis are correctly

translated into specifications, drawings, procedures, and instructions.

Contrary to the above, on March 5, 2008, the licensee identified that a 4-foot

section of suction piping within containment spray system, Train A was

approximately 50 percent voided. Voiding within the containment spray

system was due to a design deficiency that did not allow for a proper fill and

vent of the system. This was entered in the licensees corrective action

program as CAR 200803462. This finding is greater than minor because it is

similar to the Example 3j in Manual Chapter 0612, Appendix E, "Examples of

- 39 - Enclosure 2

Minor Issues," in that the presence of air within the containment spray system

suction header resulted in a condition where there was reasonable doubt on

the operability of the system. This finding is of very low safety significance

because it was a design or qualification deficiency confirmed not to result in

loss of operability.

• 10 CFR Part 50, Appendix B, Criterion III, requires measures be established

to assure that applicable regulatory requirements and design basis be

correctly translated into specifications, drawings, procedures, and

instructions. Technical Specifications 3.5.2 and 3.6.6 require that residual

heat removal and containment spray system components remain operable.

Contrary to this, measures were not adequate to assure installed center tube

diameters for the containment recirculation sump modification were correctly

accounted for by an accurate net positive suction head calculation.

The vendor supplying AmerenUE the containment recirculation sump strainer

identified that associated Vendor Calculation TDI-6002-05 for clean strainer

head loss did not account for the installed orifices located in the strainer

support plate. The size of the orifice beneath each strainer was smaller than

assumed in head loss calculations and was not large enough to prevent head

loss in excess of the net positive suction head required as defined in the

purchase specification supplied to the strainer vendor. The additional head

loss due to the calculation translation error was 2.28 feet. This resulted in

required net positive suction head being less than available. AmerenUE

performed three separate operability determination reviews to demonstrate

that the head loss margin could be recovered. The initial operability

determination on January 22, 2008, addressed the smaller support plate

orifice holes by using a separate vendor's flow analysis of the residual heat

removal and containment spray piping systems to demonstrate lower flow

and head losses than described in the FSAR. This operability determination

resulted in the limiting case flow path being the hot leg recirculation flow path.

Another operability review on March 12, 2008, addressed a nonconservative

temperature correction through the orifices. Subsequent to this, the licensee

informed the NRC that the additional nonconservative inputs were used in

the January 22, 2008, flow re-analysis of the residual heat removal system.

Additional analyses were performed to regain margin. This resulted in the

limiting case flow path changing from hot leg recirculation to cold leg

recirculation.

This example of inadequate design control was captured in the licensees

corrective action program as CARs 200800461 and 200802618. These

corrective action reviews documented three causes related to the following

design error:

• Time pressure to address Generic Letter 2004-02

• A complex design with parallel sequencing of different parts of the

design

• AmerenUE not independently verifying the vendor's design due to

perceived expertise and an approved 10 CFR Part 50, Appendix B,

- 40 - Enclosure 2

Quality Assurance program. AmerenUE did not perform a review of

the design, nor did they contract to have a third party engineering

review of the design.

This finding is greater than minor because it is similar to the Example 3j in

Manual Chapter 0612, Appendix E, Examples of Minor Issues, in that the

contractor error translating the design to the calculations resulted in a

condition where there was reasonable doubt on the operability of the ECCS.

This finding is of very low safety significance because it was a design or

qualification deficiency confirmed not to result in loss of operability. This

licensee-identified violation closes out Unresolved

Item 05000483/2008002-01.

ATTACHMENT: SUPPLEMENTAL INFORMATION

- 41 - Enclosure 2

SUPPLEMENTAL INFORMATION

KEY POINTS OF CONTACT

Licensee Personnel

B. Barton, Training Manager

M. Brandes, Consulting Engineer, Nuclear Engineering - Major Modifications

K. Bruckerhoff, Supervisor, Emergency Preparedness

F. Diya, Plant Director

T. Elwood, Supervising Engineer, Licensing

R. Farnam, Manager, Radiation Protection

K. Gilliam, Supervisor, Radiation Protection

L. Graessle, Manager, Regulatory Affairs

A. Heflin, Vice President, Nuclear

T. Herrmann, Vice President, Engineering

B. Holderness, Senior Health Physicist, Environmental Services

L. Kanuckel, Manager, Quality Assurance

D. Lantz, Superintendent of Operations Training

S. Maglio, Assistant Manager, Regulatory Affairs

R. Myatt, Supervisor, Engineering

K. Mills, Manager, Engineering

D. Neterer, Manager, Nuclear Operations

T. Parker, Trainer, Radiation Protection

S. Petzel, Engineer, Regulatory Affairs

J. Pitts, Component Engineer

V. Rider, ALARA Specialist, Radiation Protection

LIST OF ITEMS OPENED AND CLOSED

Opened

05000483/2008003-05 VIO Failure to Prevent Recurrence of Voids in ECCS Cold Leg

Recirculation Piping (Section 4OA2)

Opened and Closed

05000483/2008003-01 NCV Failure to Ensure the Suitability of the Design of the

Containment Air Cooler Control Circuitry (Section 1R15)05000483/2008003-02 NCV Inadequate Surveillance Procedure Resulted in an

Inoperable ECCS (Section 1R15)05000483/2008003-03 NCV Failure to Correct a Condition Adverse to Quality for

Diesel Generator Jacket Water O-Rings (Section 1R19)05000483/2008003-04 NCV Failure to Maintain an Adequate Technical Specification

Bases Change Process (Section 1R22)

Closed

05000483/2008001-00 LER Containment Cooler Inoperability (Section 4OA3)05000483/2008002-01 URI Containment Recirculation Sump Operability

(Section 4OA7)

A-1 Attachment

LIST OF DOCUMENTS REVIEWED

The following is a partial list of documents reviewed during the inspection. Inclusion on this list

does not imply that the NRC inspector reviewed the documents in their entirety, but rather that

selected sections or portions of the documents were evaluated as part of the overall inspection

effort. Inclusion of a document on this list does not imply NRC acceptance of the document or

any part of it, unless this is stated in the body of the inspection report.

Section 1R01: Adverse Weather Protection

Procedures

ODP-ZZ-00001, Operations Department - Code of Conduct, Revision 41

OSP-NB-00001, Class 1E Electrical Source Verification, Revision 032

OTN-NB-0001A, 4.16 KV Vital (Class 1E) Electrical System - A Train, Revision 12

OTN-NB-0001A, Addendum 5, NB01 Loss of Power Recovery, Revision 0

OTO-ZZ-00012, Severe Weather, Revision 10

Miscellaneous

AmerenUE Response to Generic Letter 2006-02, Grid Reliability and the Impact on Plant Risk

and the Operability of Offsite Power

Training Lesson Plan LP-01, Systems, Switchyard MD

Training Lesson Plan T61.0110.6, Systems, Switchyard MD

Section 1RO4: Equipment Alignment

Drawings

M-22AL01A, Piping and Instrumentation Diagram Auxiliary Feedwater System, Revision 33

M-22BB01(Q), Piping and Instrumentation Diagram Reactor Coolant System, Revision 30

M-22BB03A(Q), Piping and Instrumentation Diagram Reactor Coolant System, Revision 9

M-22BB03B(Q), Piping and Instrumentation Diagram Reactor Coolant System, Revision 9

M-22BB03C(Q), Piping and Instrumentation Diagram Reactor Coolant System, Revision 7

M-22BB03D(Q), Piping and Instrumentation Diagram Reactor Coolant System, Revision 7

M-22BG01(Q), Piping and Instrumentation Diagram Chemical and Volume Control System,

Revision 28

M-22BG03(Q), Piping and Instrumentation Diagram Chemical and Volume Control System,

Revision 52

A-2 Attachment

M-22EJ01(Q), Piping and Instrumentation Diagram Residual Heat Removal System,

Revision 57

M-22EM01(Q), Piping and Instrumentation Diagram High Pressure Coolant Injection System,

Revision 33

M-22EM02(Q), Piping and Instrumentation Diagram High Pressure Coolant Injection System,

Revision 19

M-22EP01(Q), Piping and Instrumentation Diagram Accumulator and Safety Injection,

Revision 16

Section 1RO5: Fire Protection

Miscellaneous

Drill Number 08-01, Evaluate Fire Brigade Response in a Radiation Area, dated March 27, 2008

Drill Critique Number 08-01, Unannounced Fire Drill, dated March 27, 2008

FSAR, Appendix 9.5B, Fire Hazard Analysis

Section 1R11: Licensed Operator Requalification Program

Procedures

OTA-RK-00024, Addendum 98D, Seismic Event, Revision 0

OTO-SG-0001, Design Basis Earthquake, Revision 13

Section 1R12: Maintenance Effectiveness

Procedures

EDP-ZZ-01128, Maintenance Rule Program, Revision 8

NUMARC 93-01, Industry Guideline for Monitoring the Effectiveness of Maintenance at Nuclear

Power Plants, Revision 3

Callaway Action Requests

200706892 200801644 200802854

Section 1R13: Maintenance Risk Assessment and Emergent Work Controls

Procedure

EDP-ZZ-01129, Callaway Plant Risk Assessment, Revision 14

Section 1R15: Operability Evaluations

Calculations

ARC-687, AFT Fathom 6.0 Output, Revision 0

A-3 Attachment

M-FL-18, LOCA and MSLB Containment Flood Level, Revision 1

WES-009-CALC-001, Wolf Creek/Callaway Post-LOCA Containment Water Level Calculation,

Revision 0

Callaway Action Requests

200800461 200802352 200803462

200802231 200802365 200804000

200802264 200802618

200802348 200803252

Drawings

E-018-00141, Sz. 5 2SP-1WD Schematic, Revision 19

E-018-00214, Wiring Diagram 2SP-1WD (Size 5), Revision 19

E-018-00273, Motor Control Center Ambient Compensated Overload Relay Heater Chart,

Revision 3

E-018-00847, Overload Relay Time Current Characteristics, Revision 4

E-018-00852, Sz. 5 2SP-1WD Schematic, Revision 11

E-018-00853, Wiring Diagram 2SP 1WD (Size 5), Revision 12

E-22NF01, Load Shedding and Emergency Load Sequencing Logic, Revision 5

E-23GN02, Schematic Diagram Containment Cooler Fans A and C, Revision 12

E-23GN02A, Schematic Diagram Containment Cooler Fans B and D, Revision 13

J-22GN02A, Containment Cooling System Containment Cooling Fans SGN01A and SGN01C,

Revision 2

J-22GN02C, Containment Cooling System Containment Cooling Fans SGN01B and SGN01D,

Revision 1

M-018-00943, 206 Ez-Flo Expansion Joint, Revision 0

M-22BG03, Piping and Instrumentation Diagram Chemical and Volume Control System,

Revision 52

M-22EM01, Piping and Instrumentation Diagram High Pressure Coolant Injection System,

Revision 33

M-23BG02, Piping Isometric CVCS-Max Charging Flow A and B Train Auxiliary Building,

Revision 12

A-4 Attachment

Procedures

ECA-0.1, Loss of All AC Power Recover Without Safety Injection Required, Revision 8

ECA-1.1, Loss of Emergency Coolant Recirculation, Revision 8

EDP-ZZ-04021, Review of Supplier Documents, Revision 5

ISF-SE-00N32, FCTNAL-NUC INSTM SOURCE RANGE N32, Revision 20

OSP-EJ-PV04A, Train A RHR and RCS Check Valve Inservice Test -IPTE, Revision 0

OSP-EJ-PV04B, Train B RHR and RCS Check Valve Inservice Test -IPTE, Revision 1

OTN-EN-00001, Containment Spray System, Revision 14

OTN-GN-00001, Containment Cooling and CRDM Cooling, Revision 1

Miscellaneous

Job 07513275 for SEN0032

Letter ULNRC-04884, Docket Number 50-483, Callaway Plant Unit 1, Union Electric Co.,

Facility Operating License NPF-30, Response to NRC Bulletin 2003-01, Potential Impact of

Debris Blockage on Emergency Sump Recirculation at Pressurized Water Reactor, dated

August 3, 2003

Letter ULNRC-05481, Docket Number 50-483, Callaway Plant Unit 1, Union Electric Co.,

Facility Operating License NPF-30 Response to Request for Additional Information, Response

to NRC Generic Letter 2004-02, Potential Impact of Debris Blockage on Emergency

Recirculation During Design Basis Accidents at Pressurized Water Reactors, dated

February 29, 2008

Garlock Sealing Technologies Customer Requested Test 206 Ez-Flo Expansion Test, dated

November 15, 2006

Section 1R18: Plant Modifications

Procedure

OTN-KA-00001, Compressed Air System, Revision 18

Drawings

E-018-00141, Sz. 5 2SP-1WD Schematic, Revision 19

E-018-00214, Wiring Diagram 2SP-1WD (Size 5), Revision 19

E-018-00852, Sz. 5 2SP-1WD Schematic, Revision 11

E-018-00853, Wiring Diagram 2SP 1WD (Size 5), Revision 12

E-22NF01, Load Shedding and Emergency Load Sequencing Logic, Revision 5

A-5 Attachment

E-23GN02, Schematic Diagram Containment Cooler Fans A and C, Revision 12

E-23GN02A, Schematic Diagram Containment Cooler Fans B and D, Revision 13

J-22GN02A, Containment Cooling System Containment Cooling Fans SGN01A and SGN01C,

Revision 2

J-22GN02C, Containment Cooling System Containment Cooling Fans SGN01B and SGN01D,

Revision 1

M-22KA01, Piping and Instrumentation Diagram Compressed Air System, Revision 016A

M-22KA06, Piping and Instrumentation Diagram Instrument Air Filter/Dryer Turbine, Building,

Revision 30A

Miscellaneous

Modification Package 08-0013, Containment Coolers DSGN01A/B/C/D Control Circuit Change,

Revision 0

Job

08003842

Section 1R19: Postmaintenance Testing

Procedures

APA-ZZ-00330, Preventative Maintenance Program, Revision 29

OTN-GN-00001, Containment Cooling and CRDM Cooling, Revision 14

Callaway Action Requests

200801270 200802810 200804164

Jobs

06524419 08001080 08002765

07006905 08002676 08003910

Drawings

E-018-00141, Sz. 5 2SP-1WD Schematic, Revision 19

E-018-00214, Wiring Diagram 2SP-1WD (Size 5), Revision 19

E-018-00853, Wiring Diagram 2SP 1WD (Size 5), Revision 12

E-018-00852, Sz. 5 2SP-1WD Schematic, Revision 11

E-22NF01, Load Shedding and Emergency Load Sequencing Logic, Revision 5

A-6 Attachment

E-23GN02, Schematic Diagram Containment Cooler Fans A and C, Revision 12

E-23GN02A, Schematic Diagram Containment Cooler Fans B and D, Revision 13

J-22GN02A, Containment Cooling System Containment Cooling Fans SGN01A and SGN01C,

Revision 2

J-22GN02C, Containment Cooling System Containment Cooling Fans SGN01B and SGN01D,

Revision 1

Miscellaneous

Modification Package 08-0013, Containment Coolers DSGN01A/B/C/D Control Circuit Change,

Revision 0

Simple Surveillance SP08-017, Containment Cooler Control Circuit Changes, dated April 16,

2008

Section 1R22: Surveillance Testing

Procedures

EDP-ZZ-04107, HVAC Pressure Boundary Control, Revision 19

FDP-ZZ-00101, Technical Specification Bases Control Program, Revision 6

OSP-GL-0001A, Auxiliary Building Train A Negative Pressure Test, Revision 6

OSP-EJ-P001A, RHR Train A inservice Test - Group A, Revision 44

OSP-EJ-V001B, Residual heat removal Train B valve inservice test, Revision 21

OSP-NE-0001B, Standby Diesel Generator B Periodic Tests, Revision 29

OSP-SA-00003, Emergency Core Cooling System Flow Path Verification and Venting,

Revision 30

Section 2OS1: Access Controls to Radiologically Significant Areas and

Section 2OS2: ALARA Planning and Controls

Callaway Action Requests

200703726 200800631 200800991

200703956 200800632 200801135

200710799 200800633 200801390

200711181 200800727 200801430

200711846 200800838 200802003

200711875 200800887 200802280

200711880 200800888 200803141

200711881 200800891 200803204

200711883 200800957 200803205

200800219 200800973 200803208

200800438 200800988

A-7 Attachment

Audits and Self-Assessments

Quality Assurance Audit of Radiation Protection AP08-001, February 28, 2008

Quality Assurance Supplemental Audit of Radwaste AP07-012, October 30, 2007

Simple Self-assessment Report SA07-RP-S06, January 9, 2008

Radiation Work Permits/ALARA Reviews

RWP 803321RESIN, Transfer Spent Resin from Primary Tank to Liner

ALARA Package 07-03120, Reinstall Bladders into the Recycle Hold Up Tanks

Other/Meetings/Training/Work Review

ALARA Simulator Class

Callaway Plant Long Range Dose and Source Term Reduction Plan, Revision 2

Hot Spot and Shielding Log

Job 08000834 Transfer Spent Resin from Primary Tank to Liner

Plant ALARA Review Committee Meeting

Procedures

APA-ZZ-00405, Special Nuclear Material Control and Accounting, Revision 20

APA-ZZ-01000, Callaway Plant Radiation Protection Program, Revision 26

APA-ZZ-01001, Callaway Plant ALARA Program, Revision 11

APA-ZZ-01106, Lock and Key Control, Revision 16

HDP-ZZ-01100, ALARA Planning and Review, Revision 6

HDP-ZZ-01200, Radiation Work Permits, Revision 9

HTP-ZZ-01203, Radiological Area Access Control, Revision 36

HTP-ZZ-06001, High Radiation/Very High Radiation Area Access, Revision 31

HTP-ZZ-06028, Radiological Controls for Pools that Contain or Store Spent Fuel, Revision 5

RTS-HC-00350, Primary Spent Resin Storage Tank Transfer to Bulk Waste Disposal Station,

Revision 3

Section 4OA1: Performance Indicator Verification

Procedure

NOD-QP-40, NRC Performance Indicator Program, Revision 2

Miscellaneous

Various Callaway Control Room Logs, dated March 2007 through March 2008

Callaway Integrated Inspection Report 05000483/2007002

A-8 Attachment

Callaway Integrated Inspection Report 05000483/2007003

Callaway Integrated Inspection Report 05000483/2007004

Callaway Integrated Inspection Report 05000483/2008002

Section 4OA2: Identification and Resolution of Problems

Inspection Findings

NCV 05000483/2005002-01

NCV 05000483/2006012-04

Callaway Action Requests

200501192 200800355 200804000

200709819 200800522 200804164

200711496 200801270 200805049

200800246 200801529 200805122

200800298 200801830 200808956

Generic Communications

NRC Information Notice 2006-21, OE Regarding Entrainment of Air into Emergency Core

Cooling and Containment Spray Systems, September 21, 2006

Generic Letter 2008-01, Managing Gas Accumulation in Emergency Core Cooling, Decay Heat

Removal , and Containment Spray Systems, January 11, 2009

Procedures

OTN-EM0001, Safety Injection System, Revision 27

OSP-SA-00003, Emergency Core Cooling System Flow Path Verification and Venting,

Revision 27

Section 4OA5: Other

Procedures

APA-ZZ-01004, Radiological Work Standards, Revision 9

HDP-ZZ-06100, Reactor Building Access, Revision 7

MDP-ZZ-S0001, Scaffolding Installation and Evaluation, Revision 22

OSP-EJ-00003, Containment Recirculation Sump Inspection, Revision 6

OSP-SA-00004, Visual Inspection of Containment for Loose Debris, Revision 19

OTS-ZZ-06032, Addendum 1, Debris and Particulate Control Devices, Revision 2

Calculations

Calculation BG-75, Impact of MP 06-0003, Replacement Containment Recirculation Sump

Strainers, and MP 06-0027, TSP Basket Relocation, Revision 0

Calculation BN-21, Impact of MP 06-0003, Replacement Containment Recirculation Sump

Strainer on BN21, Revision 0

A-9 Attachment

Calculation BN-22, Impact of MP 06-0003, Replacement Containment Recirculation Sump

Strainer on BN22, Revision 0

Calculation EJ-29, NPSH Margin for RHR Pumps at Transition to Recirculation When NPSH

Margin is at its Minimum Value, Revision 1

Calculation TDI-6002-05/TDI-6003-05, Clean Head Loss - Wolf Creek/Callaway, Revision 0

Callaway Action Request

200800461, Prompt Operability Determination for Containment Spray and Residual Heat

Removal Systems, Revision 0

Miscellaneous

Ameren/UE comments on ECI-PCI-WC-CAL-6002-6003-1001

Callaway Plant Containment Building Latent Debris Assessment Report Refuel 14 Fall 2005

EC-PCI-WC/CAL-6002/6003-1001, AES Calculation No. PCI-5304-S01 Structural Evaluation of

the Containment Sump Strainers, Revision 1

MP 06-0003-EC-PCI-WC-CAL-6002-6003- (000) AES Document No. PCI-5304-S01 Structural

Evaluation of the Containment Sump Strainers, Revision 1

NUREG/CR-6914, Vol. 4, Integrated Chemical Effects Test, Revision 0

SWE07848, Containment Coating Condition Assessment

TDI-6002/TDI-6003, Sure-Flow Suction Strainer Qualification Report and Addendums, Wolf

Creek/Callaway

ULNRC-05124, Docket Number 50-483 Callaway Plant Unit 1 Union Electric Co. Facility

Operating License NPF-30 Response To Generic Letter 2004-02, Potential Impact of Debris

Blockage On Emergency Recirculation During Design Basis Accidents At Pressurized-Water

Reactors.

ULNRC-05194, Docket Number 50-483 Callaway Plant Unit 1 Union Electric Co. Facility

Operating License NPF-30 September 1, 2005, Response To Generic Letter 2004-02, Potential

Impact of Debris Blockage On Emergency Recirculation During Design Basis Accidents At

Pressurized-Water Reactors.

ULNRC-05295, Docket Number 50-483 Callaway Plant Unit 1 Union Electric Co. Facility

Operating License NPF-30 Response Update To Generic Letter 2004-02, Potential Impact of

Debris Blockage On Emergency Recirculation During Design Basis Accidents At

Pressurized-Water Reactors.

ULNRC-05408, Docket Number 50-483 Callaway Plant Unit 1 Union Electric Co. Facility

Operating License NPF-30 Update For Response To Generic Letter 2004-02, Potential Impact

of Debris Blockage On Emergency Recirculation During Design Basis Accidents At

Pressurized-Water Reactors.

A-10 Attachment

ULNRC-05461, Docket Number 50-483 Callaway Plant Unit 1 Union Electric Co. Facility

Operating License NPF-30 Request for Extension Of Completion Date For Corrective Actions

Associated with NRC Generic Letter 2004-02, Potential Impact of Debris Blockage On

Emergency Recirculation During Design Basis Accidents At Pressurized-Water Reactors.

ULNRC-05465, Docket Number 50-483 Callaway Plant Unit 1 Union Electric Co. Facility

Operating License NPF-30 Supplement to Request for Extension of Corrective Actions

Completion Date For NRC Generic Letter 2004-02, Potential Impact of Debris Blockage On

Emergency Recirculation During Design Basis Accidents At Pressurized-Water Reactors.

WCAP-16568-P, Jet Impingement Testing to Determine the Zone of Influence (ZOI) for

BAQualified/Acceptable Coatings (Proprietary)

Wolf Creek/Callaway Comments on Calculation PCI-5304-S01, Structural Evaluation of the

Containment Sump Strainers.

Section 4OA7: Licensee-Identified Violations

Callaway Action Requests

200802618 200803462 200800461

Generic Communication

Generic Letter 2004-02, Potential Impact of Debris Blockage on Emergency Recirculation

During Design Basis Accidents at Pressurized Water Reactors, dated September 13, 2004

Calculation

TDI-6002-05

Correspondence

Amendment 180 to Facility Operating License NPF-30 from Mr. J. Donohew, Senior Project

Manager, Office of Nuclear Reactor Regulation to Mr. C. Naslund, AmerenUE

Procedure

APA-ZZ-00408, Professional Service Agreements and Nuclear Fuel Contracts, Revision 12

AmerenUE Callaway Plant Nuclear Plant Operating Quality Assurance Manual, Section 3,

Revision 25

Audits

Quality Assurance Audit of Design Control AP08-003

Independent Technical Review Report, SEGR 08-012, Temperature Correction Factor for

Strainer Stack Orifice Head Losses

A-11 Attachment