IR 05000483/2005009

IR 05000483-05-009; 10/1/05 - 12/31/05; Callaway Plant; Integrated Resident and Regional Report of Steam Generator Replacement Activities; Access Control to Radiologically Significant Areas

ML060370536
Person / Time
Site:	Callaway
Issue date:	02/03/2006
From:	William Jones NRC/RGN-IV/DRP/RPB-B
To:	Naslund C Union Electric Co
References
IR-05-009
Download: ML060370536 (31)

Text

February 3, 2006Charles D. Naslund, Senior Vice President and Chief Nuclear Officer Union Electric Company

P.O. Box 620 Fulton, MO 65251 SUBJECT:CALLAWAY PLANT - NRC INTEGRATED REPORT 05000483/2005009

Dear Mr. Naslund:

On December 31, 2005, the NRC completed an inspection at your Callaway Plant. Theenclosed report documents the inspection findings which were discussed on January 5, 2006, with you and other members of your staff.This inspection examined activities conducted under your license as they relate to safety andcompliance with the Commission's rules and regulations and with the conditions of your license.

Within these areas, the inspection consisted of selected examination of procedures and representative records, observations of activities, and interviews with personnel.This report documents one self-revealing finding that was evaluated under the risk significancedetermination process as having very low safety significance (Green). The NRC hasdetermined that a violation was associated with this issue. The NRC is treating this violation asa noncited violation consistent with Section VI.A.1 of the NRC Enforcement Policy because ofthe very low safety significance and because the finding was entered into your corrective action program. If you contest this noncited violation, you should provide a response within 30 days of this inspection report, with the basis for your denial, to the U.S. Nuclear Regulatory Commission, ATTN.: Document Control Desk, Washington, DC 20555-0001; with copies to the Regional Administrator Region IV; the Director, Office of Enforcement, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001; and the NRC Resident Inspector at the CallawayPlant. In accordance with 10 CFR 2.390 of the NRC's "Rules of Practice," a copy of this letter, and itsenclosure, will be available electronically for public inspection in the NRC Public DocumentRoom or from the Publicly Available Records component of NRC's document system (ADAMS). ADAMS is accessible from the NRC Web site at http://www.nrc.gov/reading-rm/adams.html (thePublic Electronic Reading Room).

Union Electric Company-2-Should you have any questions concerning this inspection, we will be pleased to discuss themwith you.

Sincerely,

/RA/William B. Jones, ChiefProject Branch B Division of Reactor ProjectsDocket: 50-483License: NPF-30

Enclosure:

NRC Inspection Report 05000483/2005009 w/attachment: Supplemental Information

REGION IV Docket:50-483 License:NPF-30 Report:05000483/2005009 Licensee:Union Electric Company Facility:Callaway Plant Location:Junction Highway CC and Highway O Fulton, Missouri Dates:October 1 through December 31, 2005 Inspectors:R. Kopriva, Senior Project Engineer, Branch BC. Stancil, Project Engineer, Branch B M. Peck, Senior Resident Inspector D. Dumbacher, Resident Inspector B. Baca, Health Physicist, Plant Support Branch B. Tharakan, Health Physicist, Plant Support Branch C. Graves, Nuclear Safety Professional Development Program G. Guerra, Health Physicist, Plant Support Branch B. Henderson, Reactor Inspector, Engineering Branch 1 L. Ellershaw, ConsultantApproved By:W. B. Jones, Chief, Project Branch B Enclosure-2-

SUMMARY OF FINDINGS

IR 05000483/2005009; 10/1/05 - 12/31/05; Callaway Plant; Integrated Resident and RegionalReport of Steam Generator Replacement Activities; Access Control to Radiologically Significant

Areas.This report covered a 3-month inspection by region based reactor inspectors and residentinspectors. One Green, self-revealing, noncited violation was identified. The significance of most findings is indicated by their color (Green, White, Yellow, or Red) using Inspection Manual Chapter 0609, "Significance Determination Process." Findings for which the significance determination process does not apply may be Green or be assigned a severity level after NRC management's review. The NRC's program for overseeing the safe operation of commercialnuclear power reactors is described in NUREG-1649, "Reactor Oversight Process," Revision 3,dated July 2000.A.

Cornerstone: Occupational Radiation Safety

Green.

The inspectors reviewed a self-revealing, noncited violation of TechnicalSpecification 5.7.2 because AmerenUE failed to control a high radiation area with dose rates greater than 1.0 rem per hour. Specifically, on September 26, 2005, the reactor vessel head was moved from the head stand and placed back on the reactor vessel without the proper radiological controls in place for a high radiation area with dose rates as high as 6.0 rem per hour. A loud noise created by the falling of a locking device on the reactor head alerted radiation protection personnel that the head lift had begun prematurely. AmerenUE's immediate corrective actions were to ensure that individuals were not present in the high radiation area and to place the reactor head in a safe condition on the reactor vessel. The finding was entered into AmerenUE's corrective action program as Callaway Action Request 200507546. The failure to control a high radiation area with dose rates greater than 1.0 remper hour is a performance deficiency. The finding was greater than minor because it was associated with the Occupational Radiation Safety Cornerstone attribute of program and process and affected the cornerstone objective to ensure the adequate protection of a worker's health and safety from exposure to radiation. The finding involved the potential for a worker's unplanned or unintended dose resulting from actions contrary to Technical Specifications.

When processed through the Occupational Radiation Safety Significance Determination Process, the finding was determined to be of very low safety significance because the finding did not involve ALARA planning or work controls, there was no overexposure or substantial potential for an overexposure, and the ability to assess dose was not compromised. In addition, this finding hascrosscutting aspects associated with human perform ance bec ause poorcoordination and communication between the head lift crew and radiation protection personnel directly contributed to the finding (Section 2OS1).

Enclosure-3-

REPORT DETAILS

Plant StatusThe Callaway Plant was shut down for Refueling Outage 14 at the beginning of the inspectionperiod. Outage work included steam generator replacement and a major turbine overhaul.

AmerenUE completed the refueling outage and synchronized the generator to the grid onNovember 19, 2005. AmerenUE returned to full power operations on November 23, 2005.

AmerenUE operated the plant at full power for the remainder of the inspection period.

REACTOR SAFETY

Cornerstones: Initiating Events, Mitigating Systems, Barrier Integrity1R08Inservice Inspection Activities (71111.08)Inspection Procedure 71111.08 requires a minimum sample size of four (as identified inSections 02.01, 02.02, 02.03, and 02.04 of the procedure).Section 02.01:Performance of Nondestructive Examination Activities Other ThanSteam Generator Tube Inspections, Pressurized Water Reactor VesselUpper Head Penetration Inspections, Boric Acid Corrosion Control

a. Inspection Scope

The inspection procedure requires the review of nondestructive examination (NDE)activities consisting of two or three different types (i.e., volumetric, surface, or visual).

The inspectors witnessed the performance of 10 ultrasonic examinations (volumetric)and 3 radiographic examinations (volumetric) on seven welds. In addition, the inspectors reviewed 4 visual examination reports representing two VT-1 examinations and 2 VT-3 examinations. The table below identifies the above examinations, which were conducted using three methods and two examination types.ComponentIdentityExaminationTypeExamination MethodSafety injection systempipe-to-pipe weld2-EP-02-3066C-WDC-001, FW 1VolumetricUltrasonic(UT Reports UT-05-105 and -106)Safety injection systempipe-to-pipe weld2-EP-02-3066B-WDC-001-FW 1VolumetricUltrasonic(UT Reports UT-05-089 and -090)Residual heat removalsystem pipe-to-pipe weld2-EJ-04-3066C-WDC-002-FW 2Volumetric Ultrasonic(UT Reports UT-05-109 and -110 ComponentIdentityExaminationTypeExamination MethodEnclosure-4-Residual heat removalsystem pipe-to-pipe weld2-EJ-04-3066B-WDC-002-FW 2VolumetricUltrasonic(UT Reports UT-05-091 and -092) Residual heat removalsystem pipe-to-pipe weld2-EJ-04-3066A-WDC-001-FW 1VolumetricUltrasonic(UT Reports UT-05-072 and -073) Safety injection systempipe-to-pipe weld 3066C-WDC-002-FW 2VolumetricRadiographyMain steam line systempipe-to-pipe weld3080D-WDC-004-FW 4Volumetric Radiography Pressurizer safety reliefvalve BB 8010B VisualVisual VT-1 (Report05040-05-006)Bolting on residual heatremoval system valveEJVH 8701AVisualVisual VT-1(Report 05040-05-

005)Essential service waterpipe supportEF04C005/1312VisualVisual VT-3(Report 05042-05-

004)Component coolingwater pipe supportEG06H010113VisualVisual VT-3(Report 05042-05-

005)For each of the NDE activities reviewed, the inspectors verified that the examinationswere performed in accordance with the specific site procedures and the applicable American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code) requirements.During review of each examination, the inspectors verified that appropriate NDEprocedures were used, that examinations and conditions were as specified in the procedure, and that test instrumentation or equipment was properly calibrated and withinthe allowable calibration period. The inspectors also verified the NDE certifications of the personnel who performed the above ultrasonic and radiographic examinations.

Finally, the inspectors observed that indications identified during the ultrasonic and radiographic examinations were dispositioned in accordance with the ASME qualified NDE procedures used to perform the examinations.The inspection procedure required review of one or two examinations with recordableindications that were accepted for continued service, to ensure that the disposition wasmade in accordance with the ASME Code. The inspectors reviewed Corrective Action

-5-Report 200403580, which documented the identification of flaws in cold leg inlet nozzlesafe end-to-elbow Weld 2-BB-01-F302. The flaws were identified and documented on April 30, 2004, during the volumetric examinations conducted during Refueling Outage 13. AmerenUE contracted to have an analytical evaluation performed by Westinghouse and Structural Integrity Associates in accordance with ASME Code Section XI, IWB-3640, which provides the specific rules for the performance of such evaluations. The results of the evaluation supported continued unit operation for3-years leaving the flaws as-is (i.e., no repairs required at this time).

The inspection procedure further required verification of one to three welds on Class 1or 2 pressure boundary piping to ensure that the welding process and welding examinations were performed in accordance with the ASME Code. The inspectors observed portions of welding performed on the two weld joints identified in the table below. Component and ClassWeld IdentityWelding ProcessReactor coolant system hotleg elbow to nozzle, Class 1Steam Generator C,3065C-WDC-001 FW 1Gas tungsten arc welding(machine)Reactor coolant systemcold leg elbow to nozzle, Class 1Steam Generator C, 3065C-WDC-002 FW 2Gas tungsten arc welding (machine)The inspectors verified, by review, that the welding procedure specifications and thewelders had been properly qualified in accordance with ASME Code,Section IX, requirements. The inspectors also verified, through observation and record review, that essential variables for the gas tungsten arc welding process and the shielded metal arc welding process had been identified, recorded in the procedure qualification record, and formed the bases for qualification of the welding procedure specifications. The inspectors also verified that weld filler materials were properly stored and controlled(there were two welding material issue areas) and that proper administrative controls were being implemented with respect to issuance and return of weld filler materials.The inspectors completed one sample under this section.

b. Findings

No findings of significance were identified.

Section 02.02: Reactor Vessel Upper Head Penetration Inspection Activities The inspection procedure requires this section to be performed after completion ofTemporary Instruction TI 2515/150. The TI had not been completed at the time of this inspection, thus this section was not performed.

-6-Section 02.03:Boric Acid Corrosion Control Inspection Activities (Pressurized WaterReactors)

a. Inspection Scope

The inspectors evaluated the implementation of AmerenUE's boric acid corrosion controlprogram for monitoring degradation of those systems that could be deleteriouslyaffected by boric acid corrosion.The inspection procedure requires review of a sample of boric acid corrosion controlwalkdown visual examination activities through either direct observation or record review. The inspectors reviewed the documentation associated with AmerenUE's boric acid corrosion control walkdown as specified in Procedure EDP-ZZ-01004, "Boric Acid Corrosion Control Program," Revision 003. Visual records of the components and equipment were also reviewed by the inspectors. Additionally, the inspectors independently performed examinations of piping containingboric acid during a walkdown of the containment building, the auxiliary building, and thesafeguards pump rooms. The inspection procedure requires verification that visual inspections emphasizelocations where boric acid leaks can cause degradation of safety significant components. The inspectors verified through direct observation and program/record review that AmerenUE's boric acid corrosion control inspection efforts are directed toward locations where boric acid leaks can cause degradation of safety-related components. The inspection procedure requires both a review of one to three engineering evaluationsperformed for boric acid leaks found on reactor coolant system piping and componentsand one to three corrective actions performed for identified boric acid leaks. The inspectors reviewed engineering evaluations associated with Corrective Action Reports 200506558, 200507149, 200507404, 200507416, and 200507245, which addressed boric acid leaks identified on residual heat removal system boltedconnections, pumps, valve stems on three safety injection accumulator valves, and a safety injection accumulator check valve. The evaluations appropriately addressed the causes and corrective actions. The inspectors completed one sample.

b. Findings

No findings of significance were identified.

-7-Section 02.04:Steam Generator Tube Inspection Activities

a. Inspection Scope

The inspection procedure requires this section to be performed on existing steamgenerators. However, AmerenUE replaced the four steam generators during this outage, and the inspections required by the inspection procedure were not performed.

Section 02.05:Identification and Resolution of Problems

a. Inspection scope

The inspection procedure requires review of a sample of problems associated with inservice inspections documented by AmerenUE in the corrective action program for appropriateness of the corrective actions.The inspectors reviewed 12 corrective action reports which dealt with inserviceinspection activities and found the corrective actions were appropriate. From this review the inspectors concluded that AmerenUE had an appropriate threshold for entering issues into the corrective action program and has procedures that direct a root cause evaluation when necessary. AmerenUE also had an effective program for applying industry operating experience.

b. Findings

No findings of significance were identified.

1R13 Maintenance Risk Assessments and Emergent Work Evaluation

a. Inspection Scope

The inspectors reviewed two risk evaluations and overall plant configuration controls forselected activities to verify compliance with Procedures EDP-ZZ-01129, "Callaway Plant Risk Assessment, Revision 8, and APA-ZZ-00150, "Outage Preparation and Execution, Revision 17. The inspectors discussed emergent work issues with work control personnel and reviewed the potential risk impact of these activities to verify that the work was adequately planned, controlled, and executed.*October 3, 2005, evaluated controls and plans for the movement of heavy loadsand equipment within containment during defueled operations (i.e., movement of old and new steam generators, and associated piping.) *October 6, 2005, movement and controls of mobile crane in containment. supporting systems.The inspectors completed 2 samples.

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b. Findings

No findings of significance were identified.

1R17 Evaluation of Permanent Plant Modifications for Steam Generator Replacement(71111.17)

a. Inspection Scope

The procedure requires the review of a minimum of five permanent plant modifications. The inspectors reviewed 17 Engineering Change Notices associated with permanent plant modification Package 03-1011, Replacement Steam Generator Installation, Revision 0, including safety evaluation screenings, safety evaluations, and calculations, to verify that they were performed in accordance with plant procedures. The specific modifications reviewed are listed in the Attachment. The inspectors also reviewed the procedures governing plant modifications to evaluate the effectiveness of programs for implementing modifications to risk-significant systems, structures, and components,such that these changes did not adversely affect the design and licensing basis of the facility. As part of this review, the inspectors reviewed modification packages and safetyevaluations associated with the steam generator replacement. The inspectors interviewed the cognizant design and system engineers for the identifiedmodifications as to their understanding of the modification packages. The inspectors evaluated the effectiveness of AmerenUE's corrective action process toidentify and correct problems concerning the performance of permanent plant modifications. In this effort, the inspectors reviewed two corrective action documents and the subsequent corrective actions pertaining to licensee-identified problems and errors in the performance of permanent plant modifications. The inspectors completed one sample.

b. Findings

No findings of significance were identified.

1R19 Postmaintenance Testing (71111.19)

a. Inspection Scope

The inspectors witnessed or reviewed the results of postmaintenance testing for twomaintenance activities. In each case, the test procedures were reviewed to determine if the test adequately verified proper performance of the components affected by outage maintenance activities. The Updated Final Safety Analysis Report, Technical Specifications, and design-basis documents were also reviewed as applicable to determine the adequacy of the acceptance criteria listed in the test procedures.*Motor-driven auxiliary feedwater Pump A full flow to the steam generators test. The inspectors reviewed the pump performance to verify the minimum and

-9-maximum flow and vibration parameters met ASME code and regulatoryrequirements. This test was performed using Procedure OSP-AL-V0002 (Job 05511505 on November 13, 2005).*Motor-driven auxiliary feedwater Pump B full flow to the steam generators test. The inspectors reviewed the pump performance to verify that the minimum and maximum flow and vibration parameters met ASME code and regulatory requirements. This test was performed using Procedure OSP-AL-V0002 (Job 05511504 on November 13, 2005).The inspectors completed two samples.

b. Findings

No findings of significance were identified.

1R20 Refueling and Outage Activities (71111.20)

a. Inspection Scope

The inspectors reviewed plant conditions and observed selected refueling outageactivities associated with Refueling Outage 14 to verify that AmerenUE maintained the plant in a configuration consistent with the requirements of Technical Specifications andwith the assumptions of the outage risk assessment. For this inspection, the inspectors reviewed the following activities as they related to entering conditions necessary for performing the steam generator replacement. Coverage of the full scope of Inspection Procedure 71111.20 is documented in NRC Inspection Report 05000483/2005005. Theinspectors observed portions of the following activities:*Clearance activities*Refueling

• Monitoring of heatup and startup activities

b. Findings

No findings of significance were identified.

1R22 Surveillance Testing (71111.22)

a. Inspection Scope

The inspectors evaluated the adequacy of periodic testing of the following importantnuclear plant equipment. This review included aspects such as preconditioning, the adequacy of acceptance criteria, test frequency, procedure adherence, record keeping, the restoration of standby equipment, the effectiveness of AmerenUE's problem identification and resolution program, and test equipment accuracy, range, and calibration. The inspectors reviewed the following tests, which were performed after completing substantial replacement of reactor coolant system pressure boundarycomponents:

-10-*Steam generator narrow range level trip bistable calibrations for 12 channels perJobs 05101243 and 05101244. These were performed using Procedures ISL-AF-0LPS1, -S2, -S3, and -S4, on October 4, 2005. *Steam generator narrow- and wide-range transmitter calibrations (19 channelsreviewed). These were performed using Procedures ISL-AE-0L501, 503, 504, 551, 552, 553, 554, 517, 518, 519, 527, 528, 529, 537, 538, 539, 547, 548, and549 on November 10-11, 2005. *Reactor coolant system (RCS) temperature average loop calibrations as requiredby Technical Specification Surveillance Requirements 3.3.1-6 and 3.3.1-7. These were performed per Procedures ISL-BB-0T421 and ISL-BB-0T411 (Jobs 04100394/910 performed on October 5, 2005, and 04100393/910 performed on October 6, 2005).*RCS flow measurement and transmitter calibrations required by TechnicalSpecification Surveillance Requirement 3.4.1.4. These were performed usingProcedure ESB-BB-03015 on November 28, 2005. The inspectors completed four samples.

b. Findings

No findings of significance were identified.2.RADIATION SAFETYCornerstone: Occupational Radiation Safety [OS]

2OS1Access Control to Radiologically Significant Areas (71121.01)

a. Inspection Scope

This area was inspected to assess AmerenUE's performance in implementing physicaland administrative controls for airborne radioactivity areas, radiation areas, high radiation areas, and worker adherence to these controls. The inspector used the requirements in 10 CFR Part 20, the Technical Specifications, and AmerenUE's procedures required by Technical Specifications as criteria for determining compliance.

During the inspection, the inspector interviewed the radiation protection manager,radiation protection supervisors, and radiation workers. The inspector performed independent radiation dose rate measurements and reviewed the following items:*Performance indicator events and associated documentation packages reportedby AmerenUE in the Occupational Radiation Safety Cornerstone *Controls (surveys, posting, and barricades) of radiation, high radiation, orairborne radioactivity areas

-11-*Radiation work permit, procedure, and engineering controls, and air samplerlocations*Conformity of electronic personal dosimeter alarm setpoints with surveyindications and plant policy; workers' knowledge of required actions when their electronic personnel dosimeter noticeably malfunctions or alarms*Barrier integrity and performance of engineering controls in airborne radioactivityareas*Physical and programmatic controls for highly activated or contaminatedmaterials (nonfuel) stored within spent fuel and other storage pools*Self-assessments, audits, licensee event reports, and special reports related tothe access control program since the last inspection*Corrective action documents related to access controls

• Licensee actions in cases of repetitive deficiencies or significant individualdeficiencies*Radiation work permit briefings and worker instructions

• Adequacy of radiological controls such as, required surveys, radiation protectionjob coverage, and contamination controls during job performance*Dosimetry placement in high radiation work areas with significant dose rategradients*Controls for special areas that have the potential to become very high radiationareas during certain plant operations*Posting and locking of entrances to all accessible high dose rate - high radiationareas and very high radiation areas*Radiation worker and radiation protection technician performance with respect toradiation protection work requirements Either because the conditions did not exist or an event had not occurred, noopportunities were available to review the following items:*Adequacy of AmerenUE's internal dose assessment for any actual internalexposure greater than 50 millirem CEDE*Changes in licensee procedural controls of high dose rate-high radiation areasand very high radiation areasTherefore, the inspector completed 21 of the required 21 samples.

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b. Findings

Introduction:

The inspector reviewed a Green self-revealing noncited violation ofTechnical Specification 5.7.2 because AmerenUE failed to control a high radiation area with dose rates greater than 1.0 rem per hour.

Description:

On September 26, 2005, AmerenUE moved the reactor vessel head fromthe head stand back to the reactor vessel without the proper radiological controls in place for a high radiation area with dose rates greater than 1.0 rem per hour. Radiation Protection personnel were in the process of establishing new radiological postings, barricades, and flashing lights when they were alerted to a premature lift of the reactor vessel head by a loud noise that was created by the locking device falling off the reactorhead. After Radiation Protection personnel observed that the head lift had begun, theyimmediately proceeded to ensure that no individuals were in the high radiation area where general area dose rates were as high as 6.0 rem per hour. The lift continued and the head was moved safely to the reactor vessel. Upon review of the event, it was determined that the head lift crew notified radiation protection personnel that the headlift was going to begin and radiation protection personnel acknowledged this notification.

However, radiation protection personnel covering the job did not understand that thehead lift was going to begin immediately. Therefore, proper controls were not in place to ensure no individuals gained unauthorized access to the high radiation area.Analysis: The failure to control a high radiation area with dose rates greater than1.0 rem per hour is a performance deficiency. This finding is greater than minor because it affected the Occupational Radiation Safety Cornerstone objective to ensure adequate protection of the worker health and safety from exposure to radiation and is associated with the cornerstone attribute of program and process. The finding involved the potential for a worker's unplanned or unintended dose resulting from actions contrary to Technical Specifications. When the finding was processed through the Occupational Radiation Safety Significance Determination Process, it was determined to be a finding of very low safety significance (Green) because it was not associated with ALARA planning or work controls, there was no overexposure or substantial potential foran overexposure, and the ability to assess dose was not compromised.Enforcement: Technical Specification 5.7.2, states, in part, that individual high radiationareas with radiation levels greater than or equal to 1.0 rem per hour accessible to personnel and located within large areas such as reactor containment where no enclosure exists for purposes of locking or being continuously guarded, shall be barricaded and conspicuously posted and a flashing light shall be activated as a warning device. On September 26, 2005, radiation workers had the potential to be exposed to aradiation field greater than 1.0 rem per hour because AmerenUE failed to post, barricade, and activate a clearly visible flashing light as a warning device in the area.

Because the finding was of very low safety significance and has been entered into the corrective action program as Callaway Action Request 200507546, this violation is being treated as a noncited violation consistent with Section VI.A of the NRC EnforcementPolicy: NCV 50-483/2005009-01, Failure to Control a High Radiation Area with Dose Rates Greater than 1.0 Rem per Hour.

-13-2OS2ALARA Planning and Controls (71121.02)

a. Inspection Scope

The inspector assessed licensee performance with respect to maintaining individual andcollective radiation exposures ALARA with respect to the steam generator replacement activities. The inspector used the requirements in 10 CFR Part 20 and AmerenUE's procedures required by Technical Specifications as criteria for determining compliance.

The inspector interviewed licensee personnel and reviewed:*Eight outage work activities scheduled during the inspection period andassociated work activity exposure estimates which were likely to result in the highest personnel collective exposures*Site-specific ALARA procedures

• ALARA work activity evaluations, exposure estimates, and exposure mitigationrequirements*Intended versus actual work activity doses and the reasons for anyinconsistencies *Interfacing between operations, radiation protection, maintenance, maintenanceplanning, scheduling, and engineering groups*Person-hour estimates provided by maintenance planning and other groups tothe radiation protection group with the actual work activity time requirements *Dose rate reduction activities in work planning

• Assumptions and basis for the current annual collective exposure estimate, themethodology for estimating work activity exposures, the intended dose outcome, and the accuracy of dose rate and man-hour estimates*Method for adjusting exposure estimates, or replanning work, when unexpectedchanges in scope or emergent work were encountered*Use of engineering controls to achieve dose reductions and dose reductionbenefits afforded by shielding*Exposures of individuals from selected work groups

• Corrective action documents related to the ALARA program and follow-upactivities such as initial problem identification, characterization, and tracking *Effectiveness of self-assessment activities with respect to identifying andaddressing repetitive deficiencies or significant individual deficiencies

-14-The sample size for this inspection was counted in NRC Inspection Report 05000483/2005005.

b. Findings

No findings of significance were identified.4.OTHER ACTIVITIES (OA)4OA1Performance Indicator Verification

a. Inspection Scope

The inspector sampled licensee submittals for the performance indicators listed belowfor the period from March 2004 through September 2005. To verify the accuracy of the performance indicator data reported during that period, performance indicator definitions and guidance contained in NEI 99-02, "Regulatory Assessment Indicator Guideline,"

Revision 2, were used to verify the basis in reporting for each data element.Occupational Radiation Safety CornerstoneOccupational Exposure Control Effectiveness Performance Indicators Licensee records reviewed included corrective action documentation that identifiedoccurrences in high radiation areas with dose rates greater than 1,000 millirem per hourat 30 centimeters (as defined in Technical Specifications), very high radiation areas (as defined in 10 CFR 20.1003), and unplanned personnel exposures (as defined in NEI 99-02). Additional records reviewed included ALARA records and whole body counts of selected individual exposures. The inspector interviewed licensee personnel that were accountable for collecting and evaluating the performance indicator data. Inaddition, the inspector toured plant areas to verify that high radiation and very highradiation areas were properly controlled.Public Radiation Safety CornerstoneRadiological Effluent Technical Specification/Offsite Dose Calculation Manual Radiological Effluent Occurrences Licensee records reviewed included corrective action documentation that identifiedoccurrences for liquid or gaseous effluent releases that exceeded performance indicator thresholds and those reported to the NRC. The inspector interviewed licenseepersonnel that were accountable for collecting and evaluating the performance indicator

data.

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b. Findings

No findings of significance were identified.4OA2Identification and Resolution of Problems (71152)1.Welding and NDE Inspection (71111.08)

a. Inspection Scope

The inspectors reviewed inservice inspection-related condition reports issued during thepast year and verified that AmerenUE identified, evaluated, corrected, and trendedproblems. In this effort, the inspectors evaluated the effectiveness of AmerenUE's corrective action process, including the adequacy of the technical resolutions.The inspectors reviewed the corrective action documents issued during the currentoutage and reviewed in detail a sample of four condition reports on the steam generator welding and nondestructive testing activities. The inspectors verified that AmerenUE identified, evaluated, corrected, and trended in accordance with the programrequirements in place at the Callaway Plant.

b. Findings

No findings of significance were identified.2.Steam Generator Replacement Outage Inspection (50001)

a. Inspection Scope

The inspectors reviewed the daily condition report summaries and nonconformancereports issued during the replacement project for risk-significant issues to see that AmerenUE was properly implementing the corrective action program. The inspectors verified that AmerenUE identified, evaluated, corrected, and trended in accordance withthe program requirements in place at the Callaway Plant.

b. Findings

No findings of significance were identified.3.Heavy Load Drop Significant Investigation Review (50001)

a. Inspection Scope

The inspectors selected Callaway Action Request 200507699, Strand Wires droppedfrom Temporary Lifting Device, for a detailed review. The inspectors' assessment of this event is documented in Section 4OA5. The report was reviewed to ensure that the full extent of any issues were identified, an appropriate evaluation was performed, and appropriate corrective actions were specified, prioritized, and implemented.

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b. Observations and Findings

There were no findings identified associated with the root cause analysis or correctiveactions specified and implemented. The inspectors observed that immediate corrective action taken by AmerenUE was to suspend all heavy load lifts in containment until adequate review could be accomplished to identify and correct the cause of this event.

Further, the inspectors observed that adequate corrective actions were implemented for subsequent heavy load lifts to preclude repetition of the event; therefore, no violation of regulatory requirements or findings were identified.4OA5Other Activities (IP 50001, Steam Generator Replacement Inspection)1.Steam Generator Removal and Replacement InspectionsPart a. in Section 02.03, "Steam Generator Removal and Replacement Inspections," ofIP 50001 requires review and inspection of the following four welding and NDE activities: Section 02.03.a.1:Where applicable, special procedures for welding and NDE Section 02.03.a.2:Training and qualifications for personnel performing welding and NDESection 02.03.a.3:NDE including radiography results and work packages for selectedweldsSection 02.03.a.4:Completion of pre-service NDE requirements for welds andcompletion of baseline eddy-current examination of new steam generator tubes

a. Inspection Scope

For each of the welding and NDE activities observed and reviewed, the inspectorsverified that the specific welding procedure specifications and NDE procedures met the applicable ASME Code requirements.The inspectors verified, by review, that the welding procedure specifications and thewelders had been properly qualified in accordance with ASME Code,Section IX, requirements. The inspectors also verified, through observation and record review, that essential variables for the gas tungsten arc welding process (machine and manual) and the shielded metal arc welding process had been identified, recorded in the procedure qualification record, and formed the bases for qualification of the welding procedure specificationsDuring review of each NDE, the inspectors verified that appropriate NDE procedureswere used, examinations and conditions were as specified in the procedure, and test instrumentation or equipment was properly calibrated and within the allowable calibration period. The inspectors also verified the NDE certifications of the personnel who performed the observed NDEs.

-17-In addition to the two observed welds identified in Section 1R08, the inspectorsobserved portions of welding on the following eight welds:Feedwater system pipe-to-pipe welds, Class 2Steam Generators A and B,3085A/B-WDC-004-FW 4Shielded metal arc weldingMain steam line systempipe-to-pipe welds, Class 2 Steam Generators A, B,and C, 3080A/B/C-WDC-002-FW 2 Shielded metal arc welding Main steam line systempipe-to-pipe welds, Class 2 Steam Generators A, B,and C, 3080A/B/C- WDC-004-FW 4 Gas tungsten arc welding(root pass - manual) and shielded metal arc welding (balance)In addition to the NDE identified in Section 1R08 above, the following NDE (includingradiography and baseline preservice examinations) were observed by the inspectors:Reactor coolant system hot legelbow-to-nozzle weldSteam Generator D, 3065D-WDC-001-FW 1RadiographyReactor coolant system cold legelbow-to-nozzle weldSteam Generator D, 3065D-WDC-002-FW 2RadiographyMain steam line system pipe-to-pipeweld3080D-WDC-002-FW 2RadiographyMain steam line system pipe-to-pipeweld3080D-WDC-004-FW 4Radiographyand ultrasonicReactor coolant system cold legelbow-to-nozzle weldSteam Generator A, 3065A-WDC-002-FW 2RadiographyReactor coolant system hot legelbow-to-nozzle weldSteam Generator A, 3065A-WDC-001-FW 1RadiographyFeedwater system pipe-to-elbow weld3085D-WDC-006-FW 6Radiographyand ultrasonicFinally, the inspectors reviewed the baseline eddy current report "Callaway PlantReplacement Steam Generators," dated March 2005, performed by AREVA in Chalon, France, on Steam Generators A (294), B (296), C (295), and D (297). With the exception of a single tube previously plugged in Steam Generator A, the eddy currentinspection scope was identical in all steam generators (i.e., full length bobbin coil examination of all 5,872 tubes, a rotating coil examination of all top-of-tubesheet/expansion transitions (both inlet and outlet sides for a total of 11,744 tests), and a rotating coil examination of the U-bend region of all tubes of the lowest 20 rows (13240 tests).

-18-The inspectors also reviewed Engineering Information Record 51-5062980-00,"Technical Summary of Callaway Plant Replacement Steam Generator Preservice Eddy Current Inspection," dated June 13, 2005. Included in the review was AREVA's Eddy Current Procedure 54-ISI-407-00 and the Procedure Qualification Record 54-PQ-407, dated January 3, 2005. The inspectors, for consistency, also reviewed the essential variables identified in three examination technique specification sheets summarized in Section 20, "Technique Qualification," of the procedure qualification and compared themto the Electric Power Research Institute's, Appendix H, guidelines for procedure qualification and three licensee site-specific examination technique specification sheets (Numbers 1, 2, and 3).

b. Findings

No findings of significance were identified.2.PostInstallation Verification and Testing Part a. in Section 02.04, "Post-Installation Verification and Testing Inspections," ofInspection Procedure 50001 requires review and inspection of the following activities: Section 02.04.1:Containment testing, as applicable Section 02.04.2:The licensee's postinstallation inspections and verifications programand its implementationSection 02.04.3:The conduct of RCS leakage testing and review that test results Section 02.04.4:The conduct of the steam generator secondary side leakage testingand review the test resultsSection 02.04.5:Calibration and testing of instrumentation affected by steam generatorreplacementSection 02.04.6:The procedures for equipment performance testing required toconfirm the design and to establish baseline measurements and the conduct of testing

a. Inspection Scope

The inspectors performed a review of AmerenUE's postinstallation verification andtesting program to verify that modifications were completed in accordance with the design; that drawings, procedures, and training have been updated as appropriate; that postinstallation walkdowns and inspections were performed to ensure equipment was restored and temporary services were removed; that equipment cleanliness has been verified; and that changes in performance of the steam generators and associatedparameters, such as flow rates, pressures, and temperatures were appropriately included in design documents and plant procedures. The inspectors used Inspection Procedures 71111.19 and 71111.22 to verify proper postmaintenance and surveillanceactivities. The results of the inspection are documented in Sections 1R19 and 1R22.

-19-No temporary containment opening was established for the Callaway steam generatorreplacement. Containment testing, as applicable, was satisfactory.Review of the Callaway primary and secondary leakage tests were conducted inaccordance with Procedure OSP-BB-00009 and satisfied ASME code and regulatory requirements. The inspectors reviewed for completeness, accuracy, and impact on plant operationsthe following calibrations, tests, licensing commitments and procedure changes.

• Callaway Action Request (CAR) 200500109 which implemented and trackedlicensing Request OL 1248 and subsequent licensing Amendment 168. The CAR specifically identified and tracked to completion each of Callaway's departments procedure revisions needed for the replacement steam generators.

The CAR was closed on November 1, 2005. The inspectors verified actual emergency operating procedures and abnormal procedures addressing setpoints and steps for both normal and adverse containment narrow- and wide-range level. *Volian Enterprises Procedure Program (VE PROM) management tool foremergency procedure updates based on setpoint security codes on December 16, 2005.*CWY 73/19T RSG's - "Water Hammer Prevention Analysis" by Framatome ANP

b. Findings

No findings of significance were identified.3.Lifting of Heavy Loads

a. Inspection Scope

Part b. in Section 02.03, "Steam Generator Removal and Replacement Inspections," ofInspection Procedure 50001 requires review and inspection of the following activities: *Review activities associated with lifting and rigging, including preparations andprocedures for rigging and heavy lifting and any required crane and rigging inspections, testing, equipment modifications, lay-down area preparations, and training. Observation

On September 29, 2005, the Mommoet operator satisfactorily performed functionalchecks of the strand jack assembly of the temporary lifting device (TLD) outside of thecontainment prior to its placement inside the containment. The hydraulic power units were not on the TLD platform and the hydraulic hoses were removed prior to moving the TLD into the containment.

-20-On October 1, 2005, the TLD was assembled on the cavity deck and lifted up to thepolar crane girders with the main girder, support legs, work platform, and strand jack with the associated hoses attached and the strand wires installed. The hydraulic hoses to the strand jack and electrical power were then connected to the power pack (controls). This was the first time that the TLD was fully assembled and powered up at the Callaway site as a complete unit. At this point, Mammoet and the Steam Generating Team personnel cleared the cavity deck in preparation for the functional test (TLD had not been declared operable). The Mammoet operator continued the set up of the strandjack assembly which required him to short stroke the wires through the strand jack.

During this first evolution all of the strand wires slipped out of the jack and fell approximately 100 feet to the Reactor cavity deck. AmerenUE performed an event team review and identified several contributing causesto the event. These included: the hydraulic hoses had been reversed (top to bottom);

the power pack was not capable of checking for reverse connection of the hydraulic hoses; the procedures were not specific for connecting the hydraulic hoses; and the hoses were not uniquely fitted for their application. These contributing causes werecommunicated to other facilities utilizing this TLD.

b. Findings

No findings of significance were identified. 4.Security Considerations and Adverse Impact

a. Inspection Scope

The inspectors checked for potential adverse impacts to the Callaway Plant site causedby outage activities, equipment configurations, etc., in accordance with Inspection Procedure 50001. The inspectors made frequent observations of security practices to verify that AmerenUE provided appropriate support for affected vital and protected area barriers during outage activities.

b. Findings

No findings of significance were identified.5.Outage Operating ConditionsThe inspectors used Inspection Procedure 71111.20 to verify proper outage conditions. The results of the inspection are documented in Section 1R20.6.Radiation Protection ControlsAn inspection to review radiation protection controls was performed during the steamgenerator replacement outage by regional office-based specialist inspectors. The results of the inspection are documented in Sections 2OS1 and 2OS2.

-21-7.Foreign Materials ControlThe inspectors performed frequent observations of the steam generator replacementactivities to verify AmerenUE was implementing proper foreign materials controls. In particular, the inspectors observed controls related to reactor coolant system and secondary side openings.

b. Findings

No findings of significance were identified.8.Temporary ServicesThe inspectors reviewed the work package and drawings, then observed the installation,use, and removal of temporary services in the containment building during the outage. Instructions for the use and controls for construction power, acetylene, oxygen, and argon were reviewed, and the actual installation of each system was compared to theapproved system sketches.

b. Findings

No findings of significance were identified.9.Storage of Removed Steam GeneratorsThe inspectors observed the transport and storage of the steam generators to the onsitestorage facility. The radiological safety plans were reviewed.

b. Findings

No findings of significance were identified.4OA6Management MeetingsExit Meeting SummaryOn October 21, 2005, the inspector presented the access control to radiologicalsignificant areas inspection results to Mr. Charles Naslund, Senior Vice President and Chief Nuclear Officer, and other members of his staff, who acknowledged the findings. On October 28, 2005, the inspectors presented the results of the inservice inspection toMr. Naslund, Senior Vice President and Chief Nuclear Officer, and other members of licensee management. Licensee management acknowledged the inspection findings.On December 14, 2005, the inspector presented the ALARA inspection results withrespect to the steam generator replacement outage activities to Mr. A. Heflin, Vice President, and other members of his staff who acknowledged the findings.

-22-On January 5, 2006, the inspector presented the inspection results toMr. Charles Naslund, Senior Vice President and Chief Nuclear Officer, and other members of his staff, who acknowledged the findings. AmerenUE confirmed that the inspectors retained no proprietary information.ATTACHMENT:

SUPPLEMENTAL INFORMATION

A-1SUPPLEMENTAL INFORMATIONKEY POINTS OF CONTACT

Licensee Personnel

F. Barton, Shift Assistant Operations Manager, Operations

P. Cryderman, Major Projects Superintendent, Engineering

F. Diya, Manager, Engineering Services

J. Doughty, Boric Acid Corrosion Control Engineer

M. Evans, Manager, Business Operations

R. Farnham, Superintendent-Night Shift, Radiation Protection

H. Floyd, Superintendent, Maintenance Instrument and Control

J. Geyer, Superintendent-Day Shift, Radiation Protection

K. Gilliam, ALARA Coordinator, Radiation Protection

A. Heflin, Vice President

E. Henson, Specialist, Regulatory Affairs

T. Herrmann, Vice President, Engineering

D. Hollabaugh, Superintendent, Employee Concerns Program

B. Huhmann, Supervisor, Engineering

G. Hurla, Supervisor, Radiation Protection

L. Kunuckel, Manager, Quality Assurance

R. Lamb, Manager, Maintenance

P. LeRoy, Engineering Manager, SGT Engineering

S. Maglio, NESY Superintendent, Engineering

P. McKenna, Assistant Superintendent Operations

R. Miller, Supervisor, Performance Impact

B. Montgomery, Inservice Inspection Engineer

T. Moser, Manager, Plant Engineering

C. Naslund, Sr. Vice President and CNO

D. Neterer, Manager, Operations

E. Olson, Superintendent, Performance Impact

J. Patterson, On-Sift Manager, Outages

S. Petzel, Engineer, Regional Regulatory Affairs

S. Reed, Supervisor, Engineering

M. Reidmeyer, Supervisor, Regional Regulatory Affairs

R. Rice, Major Projects Supervisor, Engineering

S. Sandbothe, Superintendent, Programs

J. Small, Superintendent, Chemistry

T. Steele, Department Performance Coordinator, Radiation Protection/Chemistry

K. Young, Manager, Regulatory AffairsNRC Pers onnel

M. Peck, Senior Resident Inspector

D. Dumbacher, Resident Inspector

W. Jones, Chief, Project Branch B, Division of Reactor Projects

A-2LIST OF ITEMS OPENED, CLOSED, AND DISCUSSEDOpened and Closed50-483/2005009-01NCVFailure to Control a High Radiation Area with Dose RatesGreater than 1.0 Rem per Hour (Section 2OS1).Audits and Self-AssessmentsQuality Assurance Audit of Radiation Protection AP05-001, March 11, 2005Quality Assurance Audit of Radiation Protection AP05-010, October 5, 2005Simple Surveillance Report SP05-046, October 4, 2005

Simple Surveillance Report SP05-046, October 15, 2005

Radiation Work PermitsP541360Manual Cleaning of the Reactor Vessel Head Flange and Installation ofthe Reactor Vessel Head after Core OffloadP726347500Final Headset Preparations, Reactor Vessel Head Set, Insert ProtectiveTygon Tube, Reconnect Thermocouples, and Putting Up South Cavity

HandrailsW237643Inspect and Clean the Bottom of the Reactor Vessel in the IncoreTunnel732000SGT11Reactor Coolant System (RCS) Cutting and Welding for SteamGenerator Replacement732000SGT13Pipe End Decontamination & HRA TOPO for Steam GeneratorReplacement732000SGT22Radiography for Steam Generator ReplacementProceduresNumber TitleRevisionAPA-ZZ-01000Callaway Plant Radiation Protection Program19

APA-ZZ-01004Radiological Work Standards0

HDP-ZZ-01200Radiation Work Permits6

HDP-ZZ-01500Radiological Postings18

A-3HTP-ZZ-01203Radiological Area Access Control30HTP-ZZ-06001High Radiation Area/Very High Radiation Area Access22

HTP-ZZ-06028Radiological Controls for Pools that Contain or Store SpentFuel 4HTP-ZZ-06042Industrial Radiography3RRA-ZZ-00001NRC Performance Indicator Program1EDP-ZZ-01004Boric Acid Corrosion Control3

54-ISI-407-00PreService Eddy Current Inspection of Callaway SteamGenerator Tubing

0AUE-UT-98-1Manual Ultrasonic Examination of Ferritic Piping Welds0AUE-UT-98-2Manual Ultrasonic Examination of Austenitic Piping Welds0

QEP 20.05Welding Material Control5APA-ZZ-0014310CFR50.59 Reviews0APA -ZZ-00600Design Change Control24

EDP-ZZ-04005Design Development40Welding Procedure Specification/RevisionProcedure Qualification Record/RevisionSM/1.1-1/2GT-SM/1.1-Q6/1, UE-47/3SM/1.1-2/0GT-SM/1.1-Q6/1SM/1.8-1/0GT-SM/1.8-Q4/3GT/1.1-1/2GT-SM/1.1-Q6/1GT/1/1-2/0GT-SM/1.1-Q6/1GT/1.8-1/1GT-SM/1.8-Q4/3GTM/1.1-1/0GT-SM/1.1-Q6/1

GTM/1.3-1/055-PQ7186-01/1

GTM/8.8-1/155-PQ7032-02/2MiscellaneousTraining and testing qualification/certification packages for NDE personnel

A-4Document 51-5062980-00, Technical Summary of Callaway Plant Replacement SteamGenerator Preservice Eddy Current Inspection, March 2005Callaway Plant Inservice Inspection Program Plan, October 12, 1994

Operating Instruction 55-010053-01, Narrow Groove Gas Tungsten Arc Welding - Heavy WallStainless Steel Piping, July 6, 2005ASME Code Case N-460, Alternative Examination Coverage for Class I and Class 2 Welds,Section XI, Division 1Calculations:NumberTitleRevision86-5028109-03Callaway-1 RSG Structural Design Loading Specification 3Corrective Action Reports:CAR 200400797CAR 200403520

CAR 200403555

CAR 200403617

CAR 200403644

CAR 200403655

CAR 200403789

CAR 200403912

CAR 200404050

CAR 200404515

CAR 200404867

CAR 200406269

CAR 200406758

CAR 200407668

CAR 200501358

CAR 200502666

CAR 200502914

CAR 200502962

CAR 200502978

CAR 200503115

CAR 200503447

CAR 200503494

CAR 200503524

CAR 200503646

CAR 200503893

CAR 200503895

CAR 200504334

CAR 200504545CAR 200504565CAR 200504940

CAR 200505005

CAR 200506254

CAR 200506605

CAR 200506824

CAR 200506860

CAR 200507001

CAR 200507007

CAR 200507037

CAR 200507061

CAR 200507139

CAR 200507193

CAR 200507260

CAR 200507264

CAR 200507285

CAR 200507329

CAR 200507541

CAR 200507546

CAR 200507597

CAR 200507619

CAR 200507663

CAR 200507673

CAR 200507688

CAR 200507761

CAR 200507798

CAR 200507868

CAR 200507877CAR 200507877CAR 200507918

CAR 200507925

CAR 200507926

CAR 200507959

CAR 200507985

CAR 200507985

CAR 200508000

CAR 200508001

CAR 200508080

CAR 200508086

CAR 200508145

CAR 200508192

CAR 200508220

CAR 200508241

CAR 200508399

CAR 200508459

CAR 200508507

CAR 200508509

CAR 200508719

CAR 200508844

CAR 200508873

CAR 200508875

CAR 200508876

CAR 200508877

CAR 200508878

CAR 200508879

CAR 200509137

A-5Engineering Change Requests

NumberDateNumberDateNumberDate12004/21/0517007/06/0519208/10/0513605/04/0517107/11/0519408/12/0513806/09/0517807/29/0520208/11/0514105/10/0518308/02/0520508/15/0514905/19/0518408/02/0521508/17/0515407/19/0518608/03/06Safety Evaluations

NumberTitleRevisionMP 00-1013-1RSG Component Modification0Safety Screenings

NumberTitleRevisionMP 03-1011-1Modification for Replacement Steam Generator Installation- Primary Piping and SG Supports

0LIST OF ACRONYMSALARAas low as is reasonably achievableASME CodeAmerican Society of Mechanical Engineers Boiler and Pressure Vessel Code

CARCallaway action request

CFRCode of Federal RegulationsNDEnondestructive examination

NEINuclear Energy Institute

NRCNuclear Regulatory Commission

RCSreactor coolant system

TLDtemporary lifting device