IR 05000483/2014005

IR 05000483/2014005, September 20, 2014 through December 31, 2014, Callaway Plant, Integrated Resident and Regional Report; Maintenance Effectiveness and Reactor Vessel Head Replacement

ML15036A620
Person / Time
Site:	Callaway
Issue date:	02/04/2015
From:	O'Keefe N F NRC/RGN-IV/DRP/RPB-B
To:	Diya F Union Electric Co
O'Keefe N F
References
IR 2014005
Download: ML15036A620 (73)

Text

February 4, 2015

Mr. Fadi Diya, Senior Vice President and Chief Nuclear Officer Union Electric Company P.O. Box 620 Fulton, MO 65251

SUBJECT: CALLAWAY PLANT - NRC INTEGRATED INSPECTION REPORT 05000483/2014005

Dear Mr. Diya,

On December 31, 2014, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at your Callaway Plant. On January 7, 2015, the NRC inspectors discussed the results of this inspection with you and other members of your staff. Inspectors documented the results of this inspection in the enclosed inspection report. NRC inspectors documented one finding of very low safety significance (Green) in this report. This finding involved a violation of NRC requirements. Further, inspectors documented two licensee-identified violations which were determined to be of very low safety significance in this report. The NRC is treating these violations as non-cited violations (NCVs) consistent with Section 2.3.2.a of the NRC Enforcement Policy. If you contest the violations or significance of 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, 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 Callaway Plant. If you disagree with a cross-cutting aspect assignment in this report, you should provide a response within 30 days of the date of this inspection report, with the basis for your disagreement, to the Regional Administrator, Region IV; and the NRC resident inspector at the Callaway Plant.

In accordance with Title 10 of the Code of Federal Regulations (10 CFR) 2.390, "Public Inspections, Exemptions, Requests for Withholding," a copy of this letter, its enclosure, and your response (if any) will be available electronically for public inspection in the NRC's Public Document Room or from the Publicly Available Records (PARS) component of the NRC's Agencywide Documents Access and Management System (ADAMS). ADAMS is accessible from the NRC Web site at http://www.nrc.gov/reading-rm/adams.html (the Public Electronic Reading Room).

Sincerely,/RA/

Neil O'Keefe, Branch Chief Project Branch B Division of Reactor Projects Docket Number: 50-483 License Number: NPF-30

Enclosure:

Inspection Report 05000483/2014005 w/

Attachments:

1. Supplemental Information 2. Items Requested for the Occupational Radiation Safety Inspection at Callaway Plant, October 27-31, 2014 3. Notification of Inspection and Request for Information, January 22, 2014

SUMMARY

IR 05000483/2014005; 09/20/2014 - 12/31/2014; Callaway Plant, Integrated Resident and Regional Report; Maintenance Effectiveness and Reactor Vessel Head Replacement. The inspection activities described in this report were performed between September 20 and December 31, 2014, by the resident inspectors at the Callaway Plant and inspectors from the NRC's Region IV office. One finding of very low safety significance (Green) is documented in this report. This finding involved a violation of NRC requirements. Additionally, NRC inspectors documented two licensee-identified violations of very low safety significance. The significance of inspection findings is indicated by their color (Green, White, Yellow, or Red), which is determined using Inspection Manual Chapter 0609, "Significance Determination Process." Their cross-cutting aspects are determined using Inspection Manual Chapter 0310, "Aspects within the Cross-Cutting Areas." Violations of NRC requirements are dispositioned in accordance with the NRC Enforcement Policy. NRC's program for overseeing the safe operation of commercial nuclear power reactors is described in NUREG-1649, "Reactor Oversight Process."

Cornerstone: Mitigating Systems

Green.

Inspectors reviewed a self-revealing non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, "Corrective Actions," involving the licensee's failure to correct a condition adverse to quality. Specifically, an extent of condition review for a failed gasket identified that other safety related gaskets installed in the plant were potentially less reliable; however, no action was taken. One of these gaskets failed on October 17, 2014, and caused a 52 gpm leak from the end bell of the train B control room air conditioning chiller affecting the reliability of the train. This issue was entered into the licensee's corrective action program as Callaway Action Request 201409335. One remaining improperly installed gasket was evaluated and operability was justified. The inspectors determined the failure to correct a condition adverse to quality was a performance deficiency. This performance deficiency was more than minor because it was associated with and adversely affected the equipment performance attribute of the Mitigating Systems Cornerstone, and affected the cornerstone objective to ensure the reliability of systems that respond to initiating events to prevent undesirable consequences. Specifically, during the period of November 2013 to October 2014, the failure to correct an improperly installed gasket on a control room air conditioning unit resulted in reduced reliability of the safety related system. Using NRC Inspection Manual 0609, Appendix A, Exhibit 2, "Mitigating Systems Screening Questions," the finding was determined to be of very low safety significance (Green) because it did not affect the design or qualification of the system, did not result in a loss of system function, did not represent a loss of function of a single train for greater than its technical specifications allowed outage time, and did not cause the loss of function of one or more non-technical specification trains of equipment designated as high safety-significance. The finding has an Evaluation cross-cutting aspect within the problem identification and resolution area because the licensee failed to thoroughly evaluate and ensure that the resolution addressed the extent of condition commensurate with its safety significance. Specifically, the extent of condition for improperly installed safety related gaskets was identified; however, the evaluation of the degraded condition did not assess the significance and cause corrective actions to be scheduled in a timely manner commensurate with that significance [P.2]. (Section 1R12)

=

Licensee-Identified Violations===

Two violations of very low safety significance were identified by the licensee and have been reviewed by the inspectors. Corrective actions taken or planned by the licensee have been entered into the licensee's corrective action program. These violations and associated corrective action tracking numbers are listed in Section 4OA7 of this report.

PLANT STATUS

Callaway began the inspection period at 100 percent power. On October 6, 2014, the licensee began coasting down at the end of the operating cycle and on October 11 the licensee shut the plant down to start Refueling Outage 20. The reactor was restarted on November 21 and the plant was returned to full power on November 26. On December 3 the plant tripped due to an electrical fault experienced on the excitation transformer of the main generator. Callaway remained shutdown until December 5 when the reactor was restarted. The plant returned to 100 percent power on December 7. Callaway operated at full power for the remainder of the inspection period.

REPORT DETAILS

REACTOR SAFETY

Cornerstones: Initiating Events, Mitigating Systems, and Barrier Integrity

1R01 Adverse Weather Protection

.1 Readiness for Seasonal Extreme Weather Conditions

a. Inspection Scope

On November 14, 2014, the inspectors completed an inspection of the station's readiness for seasonal extreme weather conditions. The inspectors reviewed the licensee's adverse weather procedures for winter readiness and evaluated the licensee's implementation of these procedures. The inspectors verified that prior to the onset of cold weather and snow; the licensee had corrected weather-related equipment deficiencies identified during the previous winter. The inspectors selected two risk-significant systems that were required to be protected from cold weather: refueling water storage tank condensate storage tank The inspectors reviewed the licensee's procedures and design information to ensure the systems or components would remain functional when challenged by cold weather. The inspectors verified that operator actions described in the licensee's procedures were adequate to maintain readiness of these systems. The inspectors walked down portions of these systems to verify the physical condition of the adverse weather protection features. These activities constituted one sample of readiness for seasonal adverse weather, as defined in Inspection Procedure 71111.01.

b. Findings

No findings were identified.

.2 Readiness for Impending Adverse Weather Conditions

a. Inspection Scope

Since high wind conditions were forecast in the vicinity of the facility for October 1, 2014, the inspectors completed an inspection of the station's readiness for impending adverse weather conditions. The inspectors reviewed plant design features, the licensee's procedures to respond to tornadoes and high winds, and the licensee's implementation of these procedures. The inspectors evaluated operator staffing and accessibility of controls and indications for those systems required to control the plant.

These activities constituted one sample of readiness for impending adverse weather conditions, as defined in Inspection Procedure 71111.01.

b. Findings

No findings were identified.

1R04 Equipment Alignment

.1 Partial Walkdown

a. Inspection Scope

The inspectors performed partial system walk-downs of the following risk-significant systems: October 30, 2014, train A spent fuel pool cooling November 4, 2014, train A emergency diesel generator air start system November 19, 2014, train B motor-driven auxiliary feedwater pump The inspectors reviewed the licensee's procedures and system design information to determine the correct lineup for the systems. They visually verified that critical portions of the trains were correctly aligned for the existing plant configuration. These activities constituted three partial system walk-down samples as defined in Inspection Procedure 71111.04.

b. Findings

No findings were identified.

.2 Complete Walkdown

a. Inspection Scope

On November 11, 2014, the inspectors performed a complete system walk-down inspection of the cold overpressure mitigation system. The inspectors reviewed the licensee's procedures and system design information to determine the correct lineup for the existing plant configuration. The inspectors also reviewed outstanding work orders, completed work orders, open condition reports, and other open items tracked by the licensee's operations and engineering departments. The inspectors then visually verified that the system was correctly aligned for the existing plant configuration.

These activities constituted one complete system walk-down sample, as defined in Inspection Procedure 71111.04.

b. Findings

No findings were identified.

1R05 Fire Protection

Quarterly Inspection

a. Inspection Scope

The inspectors evaluated the licensee's fire protection program for operational status and material condition. The inspectors focused their inspection on four plant areas important to safety: October 8, 2014, auxiliary building level 2047, area A-19 November 17, 2014, reactor building, area RB-1 December 23, 2014, train A safety injection and centrifugal charging pumps, area A-2 December 30, 2014, auxiliary shutdown panel, areas A-28 and A-33 For each area, the inspectors evaluated the fire plan against defined hazards and defense-in-depth features in the licensee's fire protection program. The inspectors evaluated control of transient combustibles and ignition sources, fire detection and suppression systems, manual firefighting equipment and capability, passive fire protection features, and compensatory measures for degraded conditions.

These activities constituted four quarterly inspection samples, as defined in Inspection Procedure 71111.05.

b. Findings

No findings were identified.

1R06 Flood Protection Measures

a. Inspection Scope

On December 23, 2014, the inspectors completed an inspection of the station's ability to mitigate flooding due to internal causes. After reviewing the licensee's flooding analysis, the inspectors chose one plant area containing risk-significant structures, systems, and components that were susceptible to flooding:

Train A and B emergency core cooling system pump rooms The inspectors reviewed plant design features and licensee procedures for coping with internal flooding. The inspectors walked down the selected areas to inspect the design features, including the material condition of seals, drains, and flood barriers. The inspectors evaluated whether operator actions credited for flood mitigation could be successfully accomplished.

b. Findings

No findings were identified.

1R08 Inservice Inspection Activities

The activities described in subsections 1 through 4 below constitute completion of one inservice inspection sample, as defined in Inspection Procedure 71111.08.

.1 Non-destructive Examination Activities and Welding Activities

a. Inspection Scope

The inspectors directly observed the following non-destructive examinations: System Weld Identification Examination Type Feedwater 2AE05F021 Ultrasonic Steam Generator A Main Steam Isolation Valve ABHV0014 Ultrasonic Pressurizer Skirt Weld 2-TBB03-SKIRT-W-IWA Magnetic Particle Steam Generator A Main Steam Isolation Valve ABHV0014 Phased Array Ultrasonic Pressurizer Spray Nozzle Weld Overlay 2-TBB03-SPR-07000267-SWOL Phased Array Ultrasonic Pressurizer Relief Nozzle Weld Overlay 2-TBB03-REL-07000264-SWOL Phased Array Ultrasonic The inspectors reviewed records for the following non-destructive examinations: System Weld Identification Examination Type Chemical and Volume Control 2-BG-02-H009 Visual (VT-3)

System Weld Identification Examination Type Reactor Vessel Inlet Nozzle Safe End Weld 2-RV-302-121-A Ultrasonic, Eddy Current Reactor Vessel Inlet Nozzle Safe End Weld 2-RV-302-121-B Ultrasonic, Eddy Current Reactor Vessel Inlet Nozzle Safe End Weld 2-RV-302-121-B Ultrasonic, Eddy Current Reactor Vessel Inlet Nozzle Safe End Weld 2-RV-302-121-B Ultrasonic, Eddy Current During the review and observation of each examination, the inspectors observed whether activities were performed in accordance with the ASME Code requirements and applicable procedures. The inspectors also reviewed the qualifications of all non-destructive examination technicians performing the inspections to determine whether they were current.

The inspectors reviewed records for the following welding activities: System Weld Identification Weld Type Examination Type Auxiliary Feedwater 13004618/500 Gas Tungsten Arc Weld Magnetic Particle, Liquid Penetrant Residual Heat Removal 13004624/010 Gas Tungsten Arc Weld Liquid Penetrant, Radiography High Pressure Coolant Injection 13004626/010 Gas Tungsten Arc Weld Radiography Essential Service Water 14003510/407 Gas Tungsten Arc Weld Magnetic Particle The inspectors reviewed whether the welding procedure specifications and the welders had been properly qualified in accordance with ASME Code Section IX requirements. The inspectors also determined whether essential variables were identified, recorded in the procedure qualification record, and formed the bases for qualification of the welding procedure specifications.

b. Findings

No findings were identified.

.2 Vessel Upper Head Penetration Inspection Activities

a. Inspection Scope

The licensee replaced the reactor vessel upper head. All inspection activities associated with the upper head replacement are documented in Section 4OA5. Therefore, the inspectors determined this section of Inspection Procedure 71111.08 was not applicable.

b. Findings

No findings were identified.

.3 Boric Acid Corrosion Control Inspection Activities

a. Inspection Scope

The inspectors reviewed the licensee's implementation of its boric acid corrosion control program for monitoring degradation of those systems that could be adversely affected by boric acid corrosion. The inspectors reviewed the documentation associated with the licensee's boric acid corrosion control walk-down as specified in Procedure EDP-ZZ-01004, "Boric Acid Corrosion Control Program," Revision 17, and Procedure QCP-ZZ-05048, "Boric Acid Walkdown for (Reactor Coolant System) RCS Pressure Boundary," Revision 8. The inspectors reviewed whether the visual inspections emphasized locations where boric acid leaks could cause degradation of safety-significant components, and whether evaluations used corrosion rates applicable to the affected components and properly assessed the effects of corrosion-induced wastage on structural or pressure boundary integrity. The inspectors observed whether corrective actions taken were consistent with the ASME Code, and 10 CFR Part 50, Appendix B requirements.

b. Findings

No findings were identified.

.4 Steam Generator Tube Inspection Activities

a. Inspection Scope

The licensee did not perform inspections of the steam generator tubes. No inspections were required this outage. No primary side inspections were performed. Therefore, the inspectors determined this section of Inspection Procedure 71111.08 was not applicable.

b. Findings

No findings were identified.

1R11 Licensed Operator Requalification Program and Licensed Operator Performance

.1 Review of Licensed Operator Requalification

a. Inspection Scope

On November 7, 2014, the inspectors observed the plant startup simulator training for an operating crew. The inspectors assessed the performance of the operators and the evaluators' critique of their performance.

These activities constitute completion of one quarterly licensed operator requalification program sample, as defined in Inspection Procedure 71111.11.

b. Findings

No findings were identified.

.2 Review of Licensed Operator Performance

a. Inspection Scope

The inspectors observed the performance of on-shift licensed operators in the plant's main control room. At the time of the observations, the plant was in a period of heightened activity or risk. The inspectors observed the operators' performance of the following activities: September 24, 2014, train B emergency diesel generator 24-hour run October 11, 2014, plant shut down for refueling In addition, the inspectors assessed the operators' adherence to plant procedures, including Procedure ODP-ZZ-00001, "Operations Department - Code of Conduct," and other operations department policies. These activities constitute completion of one quarterly licensed operator performance sample, as defined in Inspection Procedure 71111.11.

b. Findings

No findings were identified.

1R12 Maintenance Effectiveness

a. Inspection Scope

The inspectors reviewed two instances of degraded performance or condition of safety-related structures, systems, and components: September 13, 2014, train A emergency diesel generator jacket water leak October 17, 2014, train B control room air conditioning unit gasket failure The inspectors reviewed the extent of condition of possible common cause structures, systems, and component failures and evaluated the adequacy of the licensee's corrective actions. The inspectors reviewed the licensee's work practices to evaluate whether these may have played a role in the degradation of the structures, systems, and components. The inspectors assessed the licensee's characterization of the degradation in accordance with 10 CFR 50.65 (the Maintenance Rule), and verified that the licensee was appropriately tracking degraded performance and conditions in accordance with the Maintenance Rule.

These activities constituted completion of two maintenance effectiveness samples, as defined in Inspection Procedure 71111.12.

b. Findings

Introduction.

The inspectors reviewed a Green self-revealing, non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, "Corrective Action," for the licensee's failure to correct a condition adverse to quality identified in safety related equipment installed in the plant. Specifically, following a failure of a gasket in 2013, the licensee identified that other safety related gaskets installed in the plant were installed incorrectly; however, no action was taken to correct this condition. As a result, during system testing, the train B control room air conditioning chiller unit's inlet end bell gasket extruded resulting in excess leakage.

Description.

On October 17, 2014, during the performance of train B engineered safety feature actuation system testing, a pressure transient caused by a pump start resulted in a leak on train B of the control room air conditioning system chiller end bell gasket. This leak was quantified as approximately 52 gallons per minute. The unit was isolated and Technical Specification 3.7.11, Condition A, was entered while repairs were completed. It was restored to an operable condition within the time allowed by the technical specification action statement following replacement of the gasket.

The pressure transient exhibited during the testing was determined to be higher than normal system pressure due to the system alignment, but was still within design pressure and not significantly different than that which would be expected during an actual system actuation. As a result, the licensee reviewed the maintenance history of the unit and determined that gasket leakage had occurred on five occasions on this and similar units, dating back to 2007, during or soon after post-maintenance testing. This particular gasket had been installed under Preventative Maintenance Work Instruction 13511041 in August 2013.

In November 2013, after a failure of train B safety related switchgear chiller, a similar unit to the control room air conditioning chiller, Callaway Action Request 201308726 documented that the bolts fastening the chiller end bells were being torqued to 80 ft.-lbs.; however, this value appears to have no known technical basis. After researching the required compressive stress for the type of gaskets being used for these end bells, the licensee calculated that a torque value of 35 ft.-lbs should be used and changed their maintenance procedures to reflect this new value.

In February 2014, train A of the control room air conditioning system also experienced gasket leakage approximately six days after it had been replaced and torqued using the new specification of 35 ft.-lbs. After reviewing the skill of the craft versus the specificity of the work instructions, it was determined that extra procedural enhancements should be added to provide guidance on bolt tightening, use of lubricant, and prevention of contamination during the installation process. These were added to the corrective actions already developed for Callaway Action Request 201308726.

The licensee concluded that excessive bolting torque combined with inadequate skill of the craft and maintenance procedures on the part of the electrical maintenance technicians was leading to gasket unreliability and failures. After each of these instances of gasket failure, an extent of condition evaluation was completed that identified that this condition could exist in two safety related air conditioning units and two units that are important to safety. However, no immediate corrective actions or evaluations of whether the improperly installed gaskets constituted a degraded condition occurred. It was not until the pressure transient experienced during the testing in October 2014 that one of these susceptible gaskets failed during operation. The condition of the remaining improperly installed gasket was evaluated and operability justified in Callaway Action Request 201409335.

The licensee evaluated the impact of the gasket leak experienced on October 17, 2014, on the operability of the control room air conditioning system and the essential service water system. This evaluation concluded that the amount of leakage from the extruded gasket would not have threatened operability of the essential service water system. The amount of flow to the air conditioning unit also had enough margin to justify that it would have been able to perform its safety related function during an accident scenario. Based on this evaluation, the unit was deemed to have been operable but degraded during the period from August 2013 to October 2014.

Analysis.

The failure to correct a condition adverse to quality, identified in the plant, was a performance deficiency. Title 10 of the Code of Federal Regulations, Part 50, Appendix B, Criterion XVI, "Corrective Action," requires that conditions adverse to quality are promptly identified and corrected. The licensee identified that the improperly installed safety related gaskets were subject to an increased failure rate, but failed to evaluate the impact or take prompt action to correct the condition. This performance deficiency is more than minor because it affected the equipment performance attribute of the Mitigating Systems Cornerstone objective to ensure the reliability of systems that respond to initiating events to prevent undesirable consequences. Specifically, during the period of November 2013 to October 2014, the failure to correct an improperly installed gasket on a control room air conditioning chiller unit resulted in reduced reliability of the safety related train. Using NRC Inspection Manual 0609, Appendix A, Exhibit 2, "Mitigating Systems Screening Questions," the finding was determined to be of very low safety significance because it did not affect system design, did not result in a loss of system function, did not represent a loss of function of a single train for greater than its technical specifications allowed outage time, and did not cause the loss of function of one or more non-technical specification trains of equipment designated as high safety-significance. This finding has an Evaluation cross-cutting aspect within the problem identification and resolution cross-cutting area because the licensee failed to thoroughly evaluate this issue to ensure that the resolution addressed the extent of condition commensurate with its safety significance. Specifically, the extent of condition was identified; however, no evaluation of the significance of the degraded condition was performed [P.2].

Enforcement.

Title 10 of the Code of Federal Regulations, Part 50, Appendix B, Criterion XVI, "Corrective Action," requires, in part, that conditions adverse to quality, such as deficiencies, are promptly identified and corrected. Contrary to the above, from November 13, 2013, to October 17, 2014, the licensee failed to correct a condition adverse to quality. Specifically, Callaway Plant failed to take action to correct installed safety related gaskets in two safety related air conditioning system chillers that had been identified to have been improperly installed and were subject to failure, which constituted a condition adverse to quality. The licensee replaced the ruptured gasket and performed a new extent of condition evaluation for the remaining potentially improperly installed gasket. Because this violation was determined to be of very low safety significance (Green) and was entered into the licensee's corrective action program as Callaway Action Request 201409335, this violation is being treated as a non-cited violation in accordance with Section 2.3.2.a of the NRC Enforcement Policy: NCV 05000483/2014005-01, "Failure to Correct a Condition Adverse to Quality on Safety Related Equipment."

1R13 Maintenance Risk Assessments and Emergent Work Control

a. Inspection Scope

The inspectors reviewed three risk assessments performed by the licensee prior to changes in plant configuration and the risk management actions taken by the licensee in response to elevated risk: October 21, 2014, yellow shutdown risk due to low reactor coolant system inventory October 27, 2014, yellow shutdown risk due to only one train of spent fuel pool cooling available The inspectors verified that these risk assessment were performed timely and in accordance with the requirements of 10 CFR 50.65 (the Maintenance Rule) and plant procedures. The inspectors reviewed the accuracy and completeness of the licensee's risk assessments and verified that the licensee implemented appropriate risk management actions based on the result of the assessments. Additionally, on October 29, 2014, the inspectors observed portions of one emergent work activity that had the potential to impact barrier integrity; a vendor crane malfunctioned with the old reactor vessel head suspended from the crane, Job 13004893.

The inspectors verified that the licensee appropriately developed and followed a work plan for these activities. The inspectors verified that the licensee took precautions to minimize the impact of the work activities on unaffected structures, systems, and components. These activities constitute completion of three maintenance risk assessment and emergent work control inspection samples, as defined in Inspection Procedure 71111.13.

b. Findings

No findings were identified.

1R15 Operability Determinations and Functionality Assessments

a. Inspection Scope

The inspectors reviewed three operability determinations that the licensee performed for degraded or nonconforming structures, systems, or components: October 21, 2014, operability determination of residual heat removal system due to high vibrations on a vent pipe, Callaway Action Request 201407283 October 22, 2014, operability determination of train B emergency diesel generator due to high vibrations on the train B essential service water pump, Callaway Action Request 201407444 November 3, 2014, operability determination of train B emergency diesel generator intercooler due to the space bar having one missing locknut and one loose locknut, Callaway Action Request 201407779 The inspectors reviewed the timeliness and technical adequacy of the licensee's evaluations. Where the licensee determined the degraded structures, systems, and components to be operable, the inspectors verified that the licensee's compensatory measures were appropriate to provide reasonable assurance of operability. The inspectors verified that the licensee had considered the effect of other degraded conditions on the operability of the degraded structures, systems, and components.

These activities constitute completion of three operability and functionality review samples, as defined in Inspection Procedure 71111.15.

b. Findings

No findings were identified.

1R19 Post-Maintenance Testing

a. Inspection Scope

The inspectors reviewed four post-maintenance testing activities that affected risk-significant structures, systems, or components: November 6, 2014, train B essential service water piping replacement, Job 14003781 November 6, 2014, train B emergency diesel generator maintenance outage, Job 13505272 December 1, 2014, letdown system socket weld leak repair, Job 14006129 December 4, 2014, motor-driven auxiliary feedwater pump flow control valve to steam generator D repairs, Job 14006195 The inspectors reviewed licensing- and design-basis documents for the structures, systems, and components and the maintenance and post-maintenance test procedures. The inspectors observed the performance of the post-maintenance tests to verify that the licensee performed the tests in accordance with approved procedures, satisfied the established acceptance criteria, and restored the operability of the affected structures, systems, and components. These activities constitute completion of four post-maintenance testing inspection samples, as defined in Inspection Procedure 71111.19.

b. Findings

No findings were identified.

1R20 Refueling and Other Outage Activities

.1 Refueling Outage Activities

a. Inspection Scope

During the station's refueling outage that concluded on November 22, 2014, the inspectors evaluated the licensee's outage activities. The inspectors verified that the licensee considered risk in developing and implementing the outage plan, appropriately managed personnel fatigue, and developed mitigation strategies for losses of key safety functions. This verification included the following: Review of the licensee's outage plan prior to the outage Monitoring of shut-down and cool-down activities Verification that the licensee maintained defense-in-depth during outage activities Observation and review of reduced-inventory activities Observation and review of fuel handling activities Monitoring of heat-up and startup activities These activities constitute completion of one refueling outage sample, as defined in Inspection Procedure 71111.20.

b. Findings

No findings were identified.

.2 Other Outage Activities

a. Inspection Scope

The inspectors evaluated outage activities for an unplanned outage that began on December 3, 2014, and continued through December 6. The inspectors reviewed activities to ensure that the licensee considered risk in developing, planning, and implementing the outage schedule. The inspectors also confirmed that the licensee scheduled covered workers such that the minimum days off for individuals working on outage activities were in compliance with 10 CFR 26.205(d)(4) and (5).

The outage was caused when a ground occurred in the main generator exciter transformer. This caused a lockout of the main generator that opened both of the main generator output breakers, which resulted in a trip of the main turbine and an automatic trip of the reactor.

During the outage, the inspectors observed portions of the shutdown and monitored licensee controls over the outage activities listed below.

Clearance activities, including confirmation that tags were properly hung and equipment appropriately configured to safely support the work or testing Status and configuration of systems to ensure that technical specifications were met Management of fatigue Licensee identification and resolution of problems related to outage activities Startup and ascension to full power operation and tracking of startup prerequisites These activities constitute completion of one outage activities sample, as defined in Inspection Procedure 71111.20.

1R22 Surveillance Testing

a. Inspection Scope

The inspectors observed eight risk-significant surveillance tests and reviewed test results to verify that these tests adequately demonstrated that the structures, systems, and components were capable of performing their safety functions: In-service tests: November 25, 2014, train A safety injection pump and valve testing, Job 14508849 Reactor coolant system leak detection tests: November 29, 2014, reactor coolant system inventory balance Other surveillance tests: September 24, 2014, train B emergency diesel generator 24-hour run, Job 13503949 October 8, 2014, simultaneous start of both emergency diesel generators, Job 08502992 October 11, 2014, main turbine overspeed testing, Job 13506675 October 14, 2014, containment integrated leak rate test, Job 04501245 October 17, 2014, train A and B engineered safety features actuation system testing, Job 13506547 November 21, 2014, low power physics testing, Job 13508126 The inspectors verified that these tests met technical specification requirements, that the licensee performed the tests in accordance with their procedures, and that the results of the test satisfied appropriate acceptance criteria. The inspectors verified that the licensee restored the operability of the affected structures, systems, and components following testing. These activities constitute completion of eight surveillance testing inspection samples, as defined in Inspection Procedure 71111.22.

b. Findings

No findings were identified.

Cornerstone:

Emergency Preparedness

1EP4 Emergency Action Level and Emergency Plan Changes

a. Inspection Scope

The inspectors performed an in-office review of the "Callaway Plant Radiological Emergency Response Plan," Revision 45, and Procedure EIP-ZZ-00212, "Protective Action Recommendations," Revision 26. These revisions, Changed the description of the licensee's dose assessment model from a straight-line Gaussian to a variable trajectory model Deleted evacuation time estimates from Procedure EIP-ZZ-00212 Deleted MAGNEM as the licensee's primary method of radiological assessment Implemented the Unified RASCAL dose assessment program as the licensee's primary method of radiological assessment These revisions were compared to their previous revisions, to the criteria of NUREG-0654, "Criteria for Preparation and Evaluation of Radiological Emergency Response Plans and Preparedness in Support of Nuclear Power Plants," Revision 1, and to the standards in 10 CFR 50.47(b) to determine if the revisions adequately implemented the requirements of 10 CFR 50.54(q)(3) and 50.54(q)(4). The inspectors verified that the revisions did not decrease the effectiveness of the emergency plan. This review was not documented in a safety evaluation report and did not constitute approval of licensee-generated changes; therefore, these revisions are subject to future inspection.

These activities constitute completion of two emergency action level and emergency plan changes samples as defined in Inspection Procedure 71114.04.

b. Findings

No findings were identified.

RADIATION SAFETY

Cornerstones: Public Radiation Safety and Occupational Radiation Safety

2RS1 Radiological Hazard Assessment and Exposure Controls

a. Inspection Scope

The inspectors assessed the licensee's performance in assessing the radiological hazards in the workplace associated with licensed activities. The inspectors assessed the licensee's implementation of appropriate radiation monitoring and exposure control measures for both individual and collective exposures. The inspectors walked down various portions of the plant and performed independent radiation dose rate measurements. The inspectors interviewed the radiation protection manager, radiation protection supervisors, and radiation workers. The inspectors reviewed licensee performance in the following areas: The hazard assessment program, including a review of the licensee's evaluations of changes in plant operations and radiological surveys to detect dose rates, airborne radioactivity, and surface contamination levels Instructions and notices to workers, including labeling or marking containers of radioactive material, radiation work permits, actions for electronic dosimeter alarms, and changes to radiological conditions Programs and processes for control of sealed sources and release of potentially contaminated material from the radiologically controlled area, including survey performance, instrument sensitivity, release criteria, procedural guidance, and sealed source accountability Radiological hazards control and work coverage, including the adequacy of surveys, radiation protection job coverage and contamination controls, the use of electronic dosimeters in high noise areas, dosimetry placement, airborne radioactivity monitoring, controls for highly activated or contaminated materials (non-fuel) stored within spent fuel and other storage pools, and posting and physical controls for high radiation areas and very high radiation areas Radiation worker and radiation protection technician performance with respect to radiation protection work requirements Audits, self-assessments, and corrective action documents related to radiological hazard assessment and exposure controls since the last inspection These activities constitute completion of one sample of radiological hazard assessment and exposure controls as defined in Inspection Procedure 71124.01.

b. Findings

No findings were identified.

2RS3 In-plant Airborne Radioactivity Control and Mitigation

a. Inspection Scope

The inspectors evaluated whether the licensee controlled in-plant airborne radioactivity concentrations consistent with ALARA principles and that the use of respiratory protection devices did not pose an undue risk to the wearer. During the inspection, the inspectors interviewed licensee personnel, walked down various portions of the plant, and reviewed licensee performance in the following areas: The licensee's use, when applicable, of ventilation systems as part of its engineering controls The licensee's respiratory protection program for use, storage, maintenance, and quality assurance of National Institute for Occupational Safety and Health-certified equipment, qualification and training of personnel, and user performance The licensee's capability for refilling and transporting self-contained breathing apparatus air bottles to and from the control room and operations support center during emergency conditions, status of self-contained breathing apparatus staged and ready for use in the plant and associated surveillance records, and personnel qualification and training Audits, self-assessments, and corrective action documents related to in-plant airborne radioactivity control and mitigation since the last inspection These activities constitute completion of one sample of in-plant airborne radioactivity control and mitigation as defined in Inspection Procedure 71124.03.

b. Findings

No findings were identified.

OTHER ACTIVITIES

Cornerstones: Initiating Events, Mitigating Systems, Barrier Integrity, Emergency Preparedness, Public Radiation Safety, Occupational Radiation Safety, and Security

4OA1 Performance Indicator Verification

.1 Mitigating Systems Performance Index:

High Pressure Injection Systems (MS07)

a. Inspection Scope

The inspectors reviewed the licensee's mitigating system performance index data for the period of fourth quarter 2013 through third quarter 2014 to verify the accuracy and completeness of the reported data. The inspectors used definitions and guidance contained in Nuclear Energy Institute Document 99-02, "Regulatory Assessment Performance Indicator Guideline," Revision 7, to determine the accuracy of the reported data.

These activities constituted verification of the mitigating system performance index for high pressure injection systems, as defined in Inspection Procedure 71151.

b. Findings

No findings were identified.

.2 Mitigating Systems Performance Index:

Residual Heat Removal Systems (MS09)

a. Inspection Scope

The inspectors reviewed the licensee's mitigating system performance index data for the period of fourth quarter 2013 through third quarter 2014 to verify the accuracy and completeness of the reported data. The inspectors used definitions and guidance contained in Nuclear Energy Institute Document 99-02, "Regulatory Assessment Performance Indicator Guideline," Revision 7, to determine the accuracy of the reported data. These activities constituted verification of the mitigating system performance index for residual heat removal systems, as defined in Inspection Procedure 71151.

b. Findings

No findings were identified.

.3 Occupational Exposure Control Effectiveness (OR01)

a. Inspection Scope

The inspectors verified that there were no unplanned exposures and or losses of radiological control over locked high radiation areas and very high radiation areas during the period of second quarter 2013 through third quarter 2014. The inspectors reviewed a sample of radiologically controlled area exit transactions showing exposures greater than 100 mrem. The inspectors used definitions and guidance contained in Nuclear Energy Institute Document 99-02, "Regulatory Assessment Performance Indicator Guideline," Revision 7, to determine the accuracy of the reported data. These activities constituted verification of the occupational exposure control effectiveness performance indicator as defined in Inspection Procedure 71151.

b. Findings

No findings were identified.

===.4 Radiological Effluent Technical Specifications/Offsite Dose Calculation Manual Radiological Effluent Occurrences (PR01)

a. Inspection Scope

The inspectors reviewed corrective action program records for liquid or gaseous effluent releases that occurred from second quarter 2013 through third quarter 2014, and were reported to the NRC to verify the performance indicator data. The inspectors used definitions and guidance contained in Nuclear Energy Institute Document 99-02, "Regulatory Assessment Performance Indicator Guideline," Revision 7, to determine the accuracy of the reported data. These activities constituted verification of the radiological effluent technical specification/offsite dose calculation manual radiological effluent occurrences performance indicator as defined in Inspection Procedure 71151.

b. Findings

No findings were identified.

=

.1 Routine Review

a. Inspection Scope

Throughout the inspection period, the inspectors performed daily reviews of items entered into the licensee's corrective action program and periodically attended the licensee's condition report screening meetings. The inspectors verified that licensee personnel were identifying problems at an appropriate threshold and entering these problems into the corrective action program for resolution. The inspectors verified that the licensee developed and implemented corrective actions commensurate with the significance of the problems identified. The inspectors also reviewed the licensee's problem identification and resolution activities during the performance of the other inspection activities documented in this report.

b. Findings

No findings were identified.

.2 Semiannual Trend Review

a. Inspection Scope

The inspectors reviewed the licensee's corrective action program, performance indicators, system health reports, and other documentation to identify trends that might indicate the existence of a more significant safety issue. The inspectors verified that the licensee was taking corrective actions to address identified adverse trends. These activities constitute completion of one semiannual trend review sample, as defined in Inspection Procedure 71152.

b. Observations and Assessments The inspectors noted the following negative trends: an increasing number of missed or late hourly fire watch checks and inconsistent use of written instructions by the maintenance department. The inspectors verified that the licensee was aware of these issues and these issues have been entered into the corrective action program.

c. Findings

No findings were identified.

.3 Annual Follow-up of Selected Issues

a. Inspection Scope

On November 18, 2014, the inspectors selected Callaway Action Request 201408530, which identified that, contrary to technical specifications, multiple safety injection accumulator outlet valve breakers were closed at the same time, for an in-depth follow-up. The inspectors assessed the licensee's problem identification threshold, cause analyses, extent of condition reviews and compensatory actions. The inspectors verified that the licensee appropriately prioritized the planned corrective actions and that these actions were adequate to correct the condition and prevent recurrence. A licensee-identified violation associated with this issue is described in Section 4OA7.

The inspectors also assessed the licensee's implementation of their process used to identify, document, track, and resolve operational challenges. Inspection activities included, but were not limited to, a review of the cumulative effects of the operator workarounds on system availability and the potential for improper operation of the system, for potential impacts on multiple systems, and on the ability of operators to respond to plant transients or accidents.

The inspectors performed a review of the cumulative effects of operator workarounds. The documents listed in the attachment were reviewed to accomplish the objectives of the inspection procedure. The inspectors reviewed both current and historical operational challenge records to determine whether the licensee was identifying operator challenges at an appropriate threshold, had entered them into their corrective action program, and proposed or implemented appropriate and timely corrective actions that addressed each issue. Reviews were conducted to determine if any operator challenge could increase the possibility of an initiating event, if the challenge was contrary to training, required a change from long-standing operational practices, or created the potential for inappropriate compensatory actions. Daily plant and equipment status logs, degraded instrument logs, and operator aids or tools being used to compensate for material deficiencies were also assessed to identify any potential sources of unidentified operator workarounds.

These activities constitute completion of two annual follow-up samples, which included one operator work-around sample, as defined in Inspection Procedure 71152.

b. Findings

No findings were identified.

4OA3 Follow-up of Events and Notices of Enforcement Discretion

On December 3, 2014, at 12:33 a.m., the Callaway plant tripped. A cable located inside the main generator exciter transformer shorted against the core of the transformer. This caused the main generator lockout relays to actuate and the immediate opening of the two output breakers that transmit the electricity from the main generator to the grid. Protective relays caused the main turbine to trip offline and the reactor to shut down as designed. The NRC resident inspectors responded to the plant to review plant status, communicate the event to supervision, evaluate performance of operators and mitigating systems, and ensure proper licensee actions and evaluation of event classifications. These activities constitute completion of one event follow-up sample, as defined in Inspection Procedure 71153.

4OA5 Other Activities

Reactor Vessel Head Replacement

1. Design and Planning Inspections The inspectors used the guidance in Inspection Procedure 71007, "Reactor Vessel Head Replacement Inspection," to perform the following reactor vessel head design and planning inspection activities.

Engineering and Technical Support The inspectors reviewed engineering and technical support activities performed prior to, and during, the reactor vessel head replacement outage. This review verified that selected design changes and modifications to structures, systems, and components described in the Final Safety Analysis Report for transporting the new and old reactor vessel heads were reviewed in accordance with 10 CFR 50.59. Additionally, key design aspects and modifications associated with the reactor vessel head replacement were also reviewed.

Lifting and Rigging The inspectors reviewed engineering design, modification, and analysis associated with reactor vessel head lifting and rigging activities. This included: (1) crane and rigging equipment; (2) reactor vessel head component drop analysis; (3) safe load paths; and (4) load laydown areas. Radiation Protection The inspectors reviewed radiation protection program controls, planning, and preparation in: (1) as low as is reasonably achievable (ALARA) planning; (2) dose estimates and tracking; (3) exposure and contamination controls; (4) radioactive material management; (5) radiological work plans and controls; (6) emergency contingencies; and (7) project staffing and training plans. This review was performed in conjunction with the baseline inspections conducted during Refueling Outage 20. Security Considerations The inspectors observed security controls and reviewed security plans to verify that any potential adverse impacts were minimized. The inspectors made frequent observations of security actions to verify that the licensee had implemented the appropriate controls for affected vital and protected area barriers during the reactor head replacement activities. 2. Reactor Vessel Head Fabrication Inspections at Licensee Facility The inspectors used the guidance in Inspection Procedure 71007, "Reactor Vessel Head Replacement Inspection," to perform the following reactor vessel head fabrication inspection activities. Heat Treatment The inspectors verified that the material heat treatment used to enhance the mechanical properties of the reactor vessel head material carbon, low alloy, and high alloy chromium steels was conducted per ASME code and approved vendor procedures consistent with the applicable ASME Code,Section III, requirements. Also, the inspectors reviewed the heat treatment procedures to assure that requirements associated with the following areas were met: (1) furnace atmosphere; (2) furnace temperature distribution and calibration of measuring and recording devices; (3) thermocouple installation; (4) heating and cooling rates; (5) quenching methods; and (6) record and documentation requirements. Nondestructive Examination (NDE) Inspections were conducted to ensure the manufacturing control plan included provisions for monitoring NDE to ascertain that the NDE was performed in accordance with applicable code, material specification, and contract requirements. Welding The inspectors reviewed the documentation for the weld overlay welding operations that established a layer of stainless steel cladding on the inside of the reactor vessel head to determine if it was accomplished per design. The inspectors also selected a sample of control rod drive mechanism flange-to-nozzle welds and reviewed the following items: (1) certified mill test reports of the flange, weld material rods, and control rod drive mechanism nozzles; (2) certified mill test reports for the welding material for the reactor vessel head cladding; (3) cladding weld records, weld rod material control requisitions, traceability of weld material rods, weld procedure qualification, welder qualifications, and nonconformance reports; (4) control rod drive mechanism nozzle cladding welding inspection records, weld rod material control requisitions, traceability of weld material rods, weld procedure qualification, welder qualifications, and nonconformance reports; (5) control rod drive mechanism to nozzle welding and welds inspection records, weld rod material control requisitions, traceability of weld material rods, weld procedure qualification, welder qualifications, and nonconformance reports; and (6) NDE procedures, NDE records of the welds, NDE personnel qualifications, and certification of the NDE solvents. Procedures Inspectors verified that repair procedures had been established and that these procedures were consistent with applicable ASME code, material specification, and contract requirements by verifying: (1) repair welding was conducted in accordance with procedures qualified to Section IX of the ASME code; (2) all welders had been qualified in accordance with Section IX of the ASME code; (3) records of repairs were maintained; and (4) that requirements had been established for the preparation of certified material test reports and that the records of all required examinations and tests were traceable to the procedures to which they were performed. Code Reconciliation The inspectors reviewed the required documentation, supplemental examinations, analysis, and ASME code documentation reconciliation to ensure that the original ASME code N-Stamp remains valid, and that the replacement head complies with appropriate NRC rules and industry requirements. The inspectors also ensured that the design specification was reconciled and a design report was prepared for the reconciliation of the replacement head, verifying that they were certified by professional engineers competent in ASME code requirements. Quality Assurance Program The inspectors verified that: (1) machining was carried out under a controlled system of operation; (2) a drawing/document control system was in use in the manufacturing process; and (3) part identification and traceability was maintained throughout processing and was consistent with the manufacturer's quality assurance program. Compliance Inspection The inspectors verified that the original ASME Code,Section III, data packages for the replacement reactor vessel head were supplemented by documents included in the ASME Code,Section XI, (pre-service inspection) data packages. The inspectors examined selected manufacturing and inspection records of the finished machined reactor vessel head and verified compliance with applicable documentation requirements.

3. Reactor Vessel Head Removal and Replacement Inspections The inspectors used the guidance in Inspection Procedure 71007, "Reactor Vessel Head Replacement Inspection," to perform the following reactor vessel head removal and replacement inspection activities: Lifting and Rigging The inspectors reviewed preparations and procedures for rigging and heavy lifting including crane and rigging inspections, testing, equipment modifications, laydown area preparations, and training for the following activities: Area preparation for the outside systems Lattice boom crawler crane assembly, disassembly, and operation Hydraulic gantry lift system Outside bridge and trolley transfer system Elevated cantilevered handling device installation and use Reactor vessel head lift rig and polar crane Downender/upender fixture Old reactor vessel head removal New reactor vessel head placement Transport of old reactor vessel head to storage location Major Structural Modifications The inspectors reviewed and observed portions of the one structural modification made to facilitate reactor vessel head replacement. Specifically, the inspectors reviewed the engineering documents associated with the removal of a portion of wall in containment. The change in design of the reactor head package made the portion of wall removed from containment unnecessary. The inspectors reviewed the design change associated

with this modification. Containment Access and Integrity The inspectors observed there were no modifications to the existing containment access structure or integrity to allow for the reactor vessel head to be removed and installed. The new and old reactor vessel head were moved in and out of containment using the existing equipment hatch. Outage Operating Conditions The inspectors reviewed and observed the establishment of operating conditions including: (1) defueling; (2) reactor coolant system draindown; (3) system isolation; (4) safety tagging; (5) radiation protection controls; (6) controls for excluding foreign materials in the reactor vessel; (7) verification of the suitability of reinstalled (reused) components for use; and (8) the installation, use, and removal of temporary services. Section 1R20 of this report documents additional activities that were performed during the outage.

Storage of Removed Reactor Vessel Head The inspectors reviewed the radiological safety plans and observed the transport, storage, and radiological surveys of the old reactor vessel head to its onsite storage location. The inspectors assessed licensee performance with respect to maintaining occupational individual and collective radiation exposures as low as is reasonably achievable with respect to the reactor vessel head replacement. During the inspection, the inspectors interviewed licensee personnel and reviewed licensee performance in the following areas: ALARA work activity plans and evaluations, in-progress reviews, exposure estimates, and exposure mitigation requirements Radiation worker and radiation protection technician performance during work activities in radiation areas and high radiation areas 4. Reactor Vessel Head Post Installation Verification and Testing Inspections The inspectors used the guidance in Inspection Procedure 71007, "Reactor Vessel Head Replacement Inspection," to perform the following post-installation verification and testing inspection activities. Selective inspections were performed of the following areas: (1) containment testing; (2) licensee's post-installation inspections and verifications program and its implementation; (3) reactor coolant system leakage testing and review of test results; (4) procedures required for equipment performance testing to confirm the design and to establish baseline measurements; and (5) pre-service inspection of new welds.

b. Findings

No findings were identified.

4OA6 Meetings, Including Exit Exit Meeting Summary On October 31, 2014, the inspectors presented the radiation safety inspection results to Mr. D. Neterer, Vice President, Nuclear Operations, and other members of the licensee staff.

The licensee acknowledged the issues presented. The licensee confirmed that any proprietary information reviewed by the inspectors had been returned or destroyed. On November 20, 2014, the inspectors presented the inservice inspection results to Ms. S. Banker, Senior Director of Executive Projects, and other members of the licensee staff. The licensee acknowledged the issues presented. The licensee confirmed that any proprietary information reviewed by the inspectors had been returned or destroyed.

On November 20, 2014, the inspectors presented the reactor vessel head inspection results to Ms. S. Banker, Senior Director of Executive Projects, and other members of the licensee staff. The licensee acknowledged the issues presented. The licensee confirmed that any proprietary information reviewed by the inspectors had been returned or destroyed.

On January 5, 2014, the inspectors conducted a telephonic exit meeting to present the results of the in-office inspection of changes to the licensee's emergency plan to Mr. P. McKenna, Manager, Emergency Preparedness, and other members of the licensee staff. The licensee acknowledged the issues presented. The licensee confirmed that any proprietary information reviewed by the inspectors had been returned or destroyed. On January 7, 2015, the resident inspectors presented the inspection results to Mr. F. Diya, Senior Vice President and Chief Nuclear Officer, and other members of the licensee staff. The licensee acknowledged the issues presented. The licensee confirmed that any proprietary information reviewed by the inspectors had been returned or destroyed.

4OA7 Licensee-Identified Violations The following violations of very low safety significance (Green) were identified by the licensee and are violations of NRC requirements which meet the criteria of the NRC Enforcement Policy for being dispositioned as non-cited violations.

Title 10 of the Code of Federal Regulations, Part 50, Appendix B, Criterion V, "Instructions, Procedures, and Drawings," requires, in part, that activities affecting quality be prescribed by documented instructions, procedures, or drawings, of a type appropriate to the circumstances, and shall be accomplished in accordance with these instructions, procedures, and drawings. Attachment 3 of Callaway Procedure APA-ZZ-01400, Appendix K, "Tools for Event Prevention," is a procedure that affects quality. It states that workers shall identify conditions on a jobsite such as sensitive equipment in the work area, and continues on to state the worker needs to eliminate hazards, install appropriate barriers, or develop contingencies before proceeding with the task. Contrary to the above, on September 12, 2014, the licensee failed to accomplish an activity affecting quality in accordance with procedures. Specifically, maintenance technicians failed to follow Procedure APA-ZZ-01400, Appendix K, and identify that train A emergency diesel generator jacket water heater outlet relief valve was sensitive equipment in that it was a valve that could be accidentally repositioned if working in the area. As a result, they failed to protect the valve from inadvertent operation. During maintenance, the valve was bumped out of position during maintenance work, causing the jacket water system to slowly leak and rendered the diesel inoperable for approximately 25 hours2.893519e-4 days <br />0.00694 hours <br />4.133598e-5 weeks <br />9.5125e-6 months <br /> before being identified by operations personnel. Using NRC Inspection Manual 0609, Appendix A, Exhibit 2, "Mitigating Systems Screening Questions," the finding was determined to be of very low safety significance because it did not affect the design of the system, did not result in a loss of system function, did not represent a loss of function of a single train for greater than its technical specifications allowed outage time, and did not cause the loss of function of one or more non-technical specification trains of equipment designated as high safety-significance. This violation was entered into the licensee's corrective action program as Callaway Action Request 201406128. Technical Specification 5.4.1.a states, in part, that "written procedures shall be established, implemented, and maintained covering the applicable procedures in Regulatory Guide 1.33, "Quality Assurance Program Requirements," Revision 2, Appendix A, February 1978." NRC Regulatory Guide 1.33, Appendix A, Section 1.d. requires procedures for "Procedure Adherence and Temporary Change Method."

Procedure APA-ZZ-00100, "Written Instructions Use and Adherence," Revision 32, a procedure intended to meet the above requirements, Section 4.1, "Rules for Written Instructions," states in part, 4.1.8. Perform procedures or written instructions in the following manner: Numbered or lettered steps are performed in sequence 4.1.9. Sections of a procedure or work instruction may be performed "out of sequence" or concurrently if allowed by the document. Steps within a section MUST be performed in the order written.

Additionally, for procedures identified as Continuous Use, it states in part, "4.6.2.c. Perform the step as written in the sequence specified." Contrary to the above, on November 18, 2014, the licensee failed to implement a procedure required by Technical Specification 5.4.1.a. Specifically, a licensed operator failed to comply with Sections 4.1 and 4.6 of Procedure APA-ZZ-00100 when he entered multiple sections of Procedure OSP-BB-VL006, "RCS Pressure Isolation Valves Inservice Tests - IPTE," which was not allowed by the document. The operator directed the closing of the breakers for all four safety injection accumulator isolation valves at the same time when the procedure directed testing one valve at a time and restoring after each test.

Using NRC Inspection Manual 0609, Appendix A, Exhibit 2, "Mitigating Systems Screening Questions," the finding was determined to be of very low safety significance because it did not affect the design or qualification of a system, did not result in a loss of system function, did not represent a loss of function of a single train for greater than its technical specifications allowed outage time, and did not cause the loss of function of one or more non-technical specification trains of equipment designated as high safety-significance. This violation was entered into the licensee's corrective action program as Callaway Action Request 201408530.

A1-

SUPPLEMENTAL INFORMATION

KEY POINTS OF CONTACT

Licensee Personnel

S. Abel, Director, Engineering Projects

S. Banker, Senior Director, Executive Projects

F. Bianco, Assistant Operations Manager, Shift

R. Davis, Welding Engineer, Nuclear Engineering

F. Diya, Senior Vice President and Chief Nuclear Officer

J. Doughty, ISI Program Owner

G. Forster, NDE Level III

K. Gilliam, Supervisor, Radiation Protection

C. Graham II, Health Physicist, Radiation Protection

J. Heithold, Career Engineer, Major Modifications

A. King, Health Physicist, Radiation Protection

J. Kovar, Licensing Engineer

G. Kremer, Director, Engineering Programs

P. McKenna, Manager, Emergency Preparedness

V. Miller, Supervising Health Physicist, Radiation Protection

W. Moskopf, Project Manager, Replacement Reactor Vessel Closure Head

D. Neterer, Vice President, Nuclear Operations

J. Nurrenbern, Boric Acid Corrosion Control Program Manager

S. Petzel, Licensing Engineer, Regulatory Affairs

C. Smith, Manager, Radiation Protection

D. Thompson, Health Physicist, Radiation Protection

R. Wilson, Engineer, Snubbers and Supports

T. Witt, Licensing Engineer, Regulatory Affairs

LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED

Opened and Closed

05000483/2014005-01 NCV Failure to Correct a Condition Adverse to Quality on Safety Related Equipment (Section 1R12)

Section 1R01: Adverse Weather Protection Procedures Number Title Revision

OTO-ZZ-00007, Attachment 3 Outside Operator Cold Weather Preparation 28

OTO-ZZ-00012 Severe Weather 29

Procedures

Number Title Revision

OTS-ZZ-00007 Plant Cold Weather 29

Callaway Action Requests

201407653

Jobs

13513839

14004720

14004781

14005404

Miscellaneous

Title Date Jobs Coded for Winter November 10, 2014 Cold Weather Preparations for 2014 November 10, 2014

Section 1R04: Equipment Alignment Procedures Number Title Revision

OSP-AL-V001A, Checklist 1 Train A Auxiliary Feedwater Valve Inservice Test, Auxiliary Feedwater System Restoration 32

OSP-KJ-V002A A Diesel Air Start System Restoration 7

Callaway Action Requests

201103532

201407507

201407658

201407746

201407893

201407896

201408562

Jobs

11512381

11513051

11513386

14005294

14005310

14005370

Miscellaneous

Number Title Date

Material Receipt Inspection Report for Residual Heat Removal Suction Relief Valves October 17, 2013

597453 Material Receipt Inspection Report

October 17, 2013

Miscellaneous

Number Title Date 0255275/1 Report for Incorrect Material Received October 17, 2013

Email chain from C. Wood to W. Hinchie and R. Kokoska, Receipt of

PO 597453 NWS RHR Pump Sunction [sic] Relief Valve Testing - OPEN ITEMS October 9, 2013 0255275/2 Report for Incorrect Material Received August 6, 2014

NWS Technologies, LLC Bill to Callaway December 13, 2012

597453 SR NWS Technologies - Ameren Missouri - Callaway Energy Center Testing Completed August 2013

NWS13-105 NWS Technologies - Quality Release, HR Pump Suction Relief Valves September 4, 2013

NWS Certificate of Calibration September 4, 2013

597453 SR NWS Technologies, Certificate of Conformance, NWS Traveler 12-279 September 4, 2013 12-279 NWS Technologies, Safety Valve Test Traveler 12-279 September 4, 2013

NWS-T-50, Revision 1 NWS Technologies, Test Data 12-279 September 4, 2013

NWS-QA-P13-2, Revision 0 NWS Technologies, Final Cleanliness Inspection 12-279 September 4, 2013

Email from C. Wood to J. Reese, RHR valves September 4, 2013

597453 SR NWS Technologies, Certificate of Conformance, NWS Traveler 12-280 September 4, 2013 12-280 NWS Technologies, Safety Valve Test Traveler 12-280 September 4, 2013

NWS-T-50, Revision 1 NWS Technologies, Test Data 12-280 September 4, 2013 12-280 NWS Technologies, Safety Valve Test Traveler 12-280 September 4, 2013

NWS-QA-P13-2, Revision 0 NWS Technologies, Final Cleanliness Inspection 12-280 September 4, 2013

NWS Technologies, NWS Personnel Qualifications 2013 June 4, 2013

NWS Technologies, Inspector Certification, Jason C. Gibson

NWS Technologies, Inspector Certification, David W. Jones

NWS Technologies, Inspector Certification, D. Codey Jones

Miscellaneous

Number Title Date

NWS Technologies, Inspector Certification, Thomas P. Nederostek

NWS Technologies, Inspector Certification, John F. Reese

UOPME130037 Memo from William J. Hinchie and Jason L. Bruemmer to Craig Parker,

PO 597453 for NWS to Test RHR Pump Suction Relief Valves October 3, 2013

Email from J. Bruemmer to D. Epperson,

RO 0255275 for the Relief Valves November 18, 2014

Section 1R05: Fire Protection Procedures Number Title Revision

APA-ZZ-00700 Fire Protection Program 20

APA-ZZ-00741 Control of Combustible Materials 27

Callaway Action Requests

201406961

201407339

201407374

201407909

201407975

Miscellaneous

Title Revision Fire Preplan Manual:

Auxiliary Building Elevation 2026'-0" 38 Fire Preplan Manual:

Auxiliary Building Elevation 2047'-6" 38 Fire Preplan Manual:

Reactor Building Elevation 1978'-6"

Fire Safety Analysis Calculation:

KC-108, Fire Area:

A-28, Auxiliary Shutdown Panel Section A 1 Fire Safety Analysis Calculation:

KC-111, Fire Area:

A-33, Auxiliary Shutdown Panel Section B 1

Section 1R06: Flood Protection Measures Drawings Number Title Revision

M-01EJ01 System Flow Diagram Residual Heat Removal System E

Calculations

Number Title Revision

FL-02 Flooding of Auxiliary Building Rooms 1107-1114 0

Section 1R08: Inservice Inspection Activities Procedures/Work Instructions Number Title Revision/Date

EDP-ZZ-01004 Boric Acid Corrosion Control Program 17

ESP-ZZ-01016 ASME Section XI IWE Containment Pressure Boundary Inspection 6

MDP-ZZ-LM001 Fluid Leak Management Program 15

MSM-ZZ-QW005 Mechanical Snubber Functional Test 17

QCP-ZZ-05041 Visual Examination to ASME

VT-2 26

QCP-ZZ-05048 Boric Acid Walkdown for Reactor Coolant System Pressure Boundary 8

QCP-ZZ-05049 Reactor Pressure Vessel Head Bare Metal Examination 3 08510270.500 Perform Ultrasonic Examination 3 11510665/500 Ultrasonic Examination of ABHV0014 Bonnet Bolting September 28, 2011 12503620.500 Perform Surface Exam (MT) of 2-TBB03-SKIRT-W-IWA 1 08510323/500 Ultrasonic ISI Examinations TBB03 October 1, 2008 13004618.500 Tie-In Installation of New Line#

AL-092-DBC-4",

MP 13-0024 1 13001424.010 Pre Fabrication of New Residual Heat Removal Secondary Make up Tie-In Point 1 13004626.010 Pre Fabrication of EM High Pressure Injection Point Tie-In 4 14004657.500 Clean/Repair Boron Leak at BGFT0138A 0 13002866.0 Perform Leak Detection of the Refueling Pool 1 13002866.400 Perform Refueling Pool Leak Detection Prior to Filling 2 13002866.500 Perform Leak Detection of the Refueling Pool 2 13002866.510 Engineering Support for Refuel Pool Leak Detection 0

Non-destructive Examination Reports Numbers

UT-14-007

UT-14-001

MT-14-001

UT-14-021

UT-14-022

Jobs 08510270.500 12503620.500 13004618.500 13001424.010 13004626.010 14003510/407 11510665/500 08510323/500 14003510/530 14004657.500

14005105 13002866.0

13002867 PM0917069 PM0917071 PM0917072

Weld Rod Issue Requisitions

205819

205820

205825

205826

205836

205837

205842

205843

205865

205866

205884

205885

205889

205890

205895

205896

205899

205900

205901

205902

205903

205904

205905

205906

205908

205909

205932

205933

205951

205952

205959

205960

205968

205969

205974

205975

205979

205980

205981

205982

205983

205984

205991

205992

205998

205999

206000

206001

206002

206003

206004

206005

206006

206007

206052

206065

206066

206070

206071

Callaway Action Requests

201004793

201108817

201109703

201109796

201302873

201303121

201303435

201303819

201401547

201406415

201406528

201406732

201406973

201406977

201406993

201407128

201407129

201407138

201407170

201407174

201407183

201407329

Miscellaneous

Number Title Revision/Date

NET 14-0034 Boric Acid Corrosion Control Program (BACCP) Quarterly Report October 5, 2014

NET 14-0023 Boric Acid Corrosion Control Program (BACCP) Quarterly Report July 3, 2014

Miscellaneous

Number Title Revision/Date

NET 14-0012 Boric Acid Corrosion Control Program (BACCP) Quarterly Report April 15, 2014

NET 13-0059 Boric Acid Corrosion Control Program (BACCP) Quarterly Report October 2, 2013

NET 13-0046 Boric Acid Corrosion Control Program (BACCP) Quarterly Report July 8, 2013

NET 14-0001 Boric Acid Corrosion Control Program (BACCP) Quarterly Report January 15, 2014

ULNRC-01779 Docket Number 50-483 Callaway Plant Response to Generic Letter 88-05 Boric Acid Corrosion of Carbon Steel Reactor Coolant Boundary Components in PWR Plants May 27, 1988 T62.7211 Q Callaway Engineering Qualification Standard - Perform Activities for Implementation of the Boric Acid Corrosion Control Program 4

201309684-08 Self-Assessment Topic:

Material Degradation Management Plan September 24, 2014

201309684-08 Self-Assessment Topic:

Inservice Inspection Program June 20, 2014

201100239 Benchmark Topic:

Compare Welding and ASME Section XI Repair/Replacement Administrative Procedures July 22, 2011 SA09-MM-04 Self-Assessment Topic:

Effectiveness Review

CAR 200705558, Welder not Certified for the Job July 29, 2009 SA00-NE-007 Self-Assessment ASME XI Repair/Replacement and Welding Program October 2000

WDI-PJF-1308925-FSR-001 Callaway Nuclear Power Plant Reactor Vessel Inlet Nozzle DM Weld and Clad Patch Inspection April 2013

ULNRC-05271 Docket Number 50-483 Union Electric Company Callaway Plant 10

CFR 50.55a Requests for Relief from ASME Section XI Inservice Inspection Requirements for Third 10-Year Inspection Interval

March 28, 2006

ULNRC-05292 Docket Number 50-483 Union Electric Company Callaway Plant 10

CFR 50.55a Request for Relief from ASME Section XI Repair and Replacement Requirements:

Proposed Alternative for Application of Structural Weld Overlays to Pressurizer Nozzle Welds August 14, 2006

ULNRC-05344 Docket Number 50-483 Union Electric Company Callaway Plant Response to Request for Additional Information Regarding 10

CFR 50.55a Request 13R-01 (Request for Relief from ASME Section XI Inservice Inspection Requirements for Third 10-Year Inspection Interval)

November 29, 2006

Miscellaneous

Number Title Revision/Date

ULNRC-06048 Callaway Plant Unit 1 Union Electric Co. Facility Operating License

NPF-30 10

CFR 50.55a Request:

Proposed Alternative to ASME Section XI Requirements Regarding System Leakage Tests of Class 1 Piping and Components Isolated Between Normally Closed Valves October 17, 2013

ULNRC-06071 Docket Number 50-483 Callaway Plant Unit 1 Union Electric Co. Facility Operating License

NPF-30 Response to Request for Additional Information RE:

13R-15, Proposed Alternative Regarding Pressure Retaining Boundary During System Leakage Test (TAC No. MF2921) January 29, 2014

ULNRC-06092 Docket Number 50-483 Callaway Plant Unit 1 Union Electric Co. Facility Operating License

NPF-30 Relief Request 13R-16 for Temporary Non-Code Repair of the train A Component Cooling Water Heat Exchanger March 19, 2014

ULNRC-06093 Docket Number 50-483 Callaway Plant Unit 1 Union Electric Co. Facility Operating License

NPF-30 Proposed Alternative to ASME Code,Section XI Requirements, which Extends Reactor Vessel Inservice Inspection Frequency from 10 to 20 Years (Relief Request 13R-17) April 8, 2014

Callaway Energy Center Steam Generator Tube Inspection Report 2012

Section 1R11: Licensed Operator Requalification Program Procedures Number Title Revision

OTG-ZZ-00001, Addendum 1 Auxiliary Spray Operation 5

OTG-ZZ-00002 Reactor Startup 54

OTG-ZZ-00003 Plant Startup Hot Zero Power to 30% Power - IPTE 59

OTG-ZZ-00004 Power Operations 88

OTG-ZZ-00004, Addendum 2 End of Life Coastdown Operations 6

Miscellaneous

Number Title Revision T61.JITTS Just In Time Training Session Number

20141672 0

Section 1R12: Maintenance Effectiveness Procedures Number Title Revision

APA-ZZ-00500, Appendix 14 Adverse Condition - Significance Level 3 20

APA-ZZ-01400, Appendix K Tools for Event Prevention 10

EDP-ZZ-01128, Appendix 2 Summary of SSC Performance Criteria 26

Miscellaneous

Number

Suggestion Occurrence Solution

199200357

Modification Package 00-1018

Modification Package 10-0003

Modification Package 10-0004

Section 1R13: Maintenance Risk Assessment and Emergent Work Controls Procedures Number Title Revision

APA-ZZ-00365 Callaway Lifting and Rigging Program 25

APA-ZZ-00365, Addendum L Callaway Lifting Operations 16

EDP-ZZ-01129, Attachment 5B Shutdown Safety Assessment - Mode 5 - Loops not Filled or Mode 6 - Reactor Coolant System Inventory Between 3 feet Below Vessel Flange (Indicated Level of 64.1") and 23 feet Above Vessel Flange (Indicated Level of 376.0") 42

EDP-ZZ-01129, Attachment 7 Shutdown Safety Assessment - No Mode 42

ODP-ZZ-00002, Appendix 1 Protected Equipment Program 22

ODP-ZZ-00002, Appendix 2 Postings for Lowered Inventory Operations 2

OTO-EJ-00001 Loss of Residual Heat Removal Flow 31

Callaway Action Requests

201407696

201407713

Jobs 13004893

Section 1R15: Operability Evaluations Procedures Number Title Revision

MSM-KJ-QK001 Emergency Diesel Generator Inspection 34

Callaway Action Requests

201303740

201407283

201407444

Jobs 12503092

Section 1R19: Post-Maintenance Testing Procedures Number Title Revision

MSM-KJ-QK001 Emergency Diesel Generator Inspection 34

OSP-AL-PV04B Train B Motor Driven Auxiliary Feedwater Comprehensive Pump and Check Valve Test 17

OSP-AL-PV04B, Attachment 1 Train B Valve Test Data 17

OSP-AL-V001B, Attachment 1 Valve Test Data Sheet 52

OSP-SA-00004 Visual Inspection of Containment for Loose Debris 25

Drawings

Number Title Revision M-23BG22 Piping Isometric, CVCS - Letdown, Reactor Building 16

Callaway Action Requests

201010669

201404221

201404656

201407757

201408809

Jobs

12503092

13505272

14002786

14003781

14005318

14005357

14006129

14006195

Miscellaneous

Title Date Burns & McDonnell letter to Mr. Roger Andreasen regarding Evaluation of Localized Defects in

EF-009-HBC-30" November 5, 2014

Section 1R20: Refueling and Other Outage Activities Procedures Number Title Revision

APA-ZZ-00908 Fitness For Duty Programs 32

APA-ZZ-00911 Fatigue Management 4

ETP-ZZ-00012 Inverse Count Rate Ratio Monitoring for Approach to Criticality 14

OSP-SA-00004 Visual Inspection of Containment for Loose Debris 25

OTG-ZZ-00002 Reactor Startup - IPTE 54

OTG-ZZ-00003 Plant Startup Hot Zero Power to 30 Percent Power - IPTE 59

OTG-ZZ-00004 Power Operations 88

OTG-ZZ-00005 Plant Shutdown 20 Percent Power to Hot Standby 45

OTG-ZZ-00006 Plant Cooldown Hot Standby to Cold Shutdown 72

OTG-ZZ-00007 Refueling Preparation, Performance and Recovery 35

OTN-BB-00001 Reactor Coolant System - IPTE 44

OTN-BB-00001, Addendum 1 Reactor Coolant System Vacuum Eductor Connections 2

OTN-BB-00002, Addendum 6 Draining the Reactor Coolant System to Limited Inventory or Reduced Inventory - IPTE 24

OTS-KE-00013 Refueling Machine 31

Callaway Action Requests

201407675

201407934

201407968

201408487

Jobs

13505838

13508285

Miscellaneous

Number Title Revision/Date

Callaway Energy Center RF20 Shutdown Safety Management Plan August 19, 2014 Figure 14.9 Callaway Plant Pressure and Temperature Limits Report 6

Callaway Refueling Outage 20 - Overview October 10, 2014

Refuel 20 - Major Scope Summary September 23, 2014

Refuel 20 - Operation Control Room Schedule September 26, 2014

Refuel 20 - Outage Organization October 7, 2014 32-9180957 Areva Calculation Summary Sheet:

Callaway Reactor Vessel Head Drop Model 1

Section 1R22: Surveillance Testing Procedures Number Title Revision

ESP-ZZ-00024 Low Power Physics Testing Data Acquisition 9

ETP-ZZ-ST010 Low Power Physics Test Program With Dynamic Rod Worth Measurement - IPTE 11

ODP-ZZ-00029 RCS Leakage Action Level Guideline 4

OSP-AC-00005 Turbine Actual Overspeed Trip 11

OSP-BB-00009 RCS Inventory Balance 36

OSP-EM-P001A Safety Injection Train A Inservice Test - Group B 50

OSP-NE-0001B Standby Diesel Generator B Periodic Tests 59

OSP-NE-00002 Simultaneous Start of Both Diesel Generators 20

OSP-NE-00002, Attachment 1 Simultaneous Start of Both Diesel Generators Data Sheet 20

OSP-NE-0024B Standby Diesel Generator B 24 Hour Run and Hot Restart Test 49

Callaway Action Requests

201406967

201407222

201407237

201407239

201407245

201407246

201407248

201407254

201407268

201407270

Callaway Action Requests

201407290

201407305

201407311

201408807

201408809

Jobs

04501245

08502992

13503547

13503949

13506675

13508126

14004748

14004966

14508849

Miscellaneous

Number Title Date Calculation

GN-17 Containment Cooler Water Hammer Analysis July 21, 1992

Letter RE:

Fathom/Impluse [sic] Comparison of SGK04 and SGN01B Pressures, from T. DeVincentis, Archon Engineering, P.C. to C. Norman October 23, 2014

Section 2RS1: Radiological Hazard Assessment and Exposure Controls Procedures Number Title Revision

APA-ZZ-01000 Callaway Energy Center Radiation Protection Program 39

APA-ZZ-01000 Appendix A Control of Radioactive Material 16

HDP-ZZ-01500 Radiological Postings 42

HTP-ZZ-02004 Control of Radioactive Sources 38

HTP-ZZ-06001 High Radiation/Locked High Radiation/Very High Radiation Area Access 47

Audit and Self Assessments Number Title Date

201308863-07 Self-Assessment - HRA/LHRA/VHRA Controls March 21, 2014

201309864-24 Self-Assessment - STARS Radiation Protection Fundamentals June 20, 2014

OQC 14-003 Nuclear Oversight Performance Report 4th Quarter 2013 January 27, 2014

Callaway Action Requests

200503146

201302963

201303275

201303901

201306521

201308866

201400781

201401142

201405757

201407541

Radiation Work Permits Number Title Revision

13004060 BB8948A Disassembly and Repair 1

13004098 EMV0057 Radiography 2

200812209 Replace B Reactor Coolant Pump (PBB01B) Cartridge Seal Assembly 0

Survey Documentation Number Title Date

CA-M-20140731-3 7404D Demin Valve Compartment July 31, 2014

CA-M-20140819-11 1308B Valve Compartment - Depost from HRA August 19, 2014

CA-M-20140820-3 7404A Demin Valve Compartment August 20, 2014

CA-M-20140829-3 7404B Demin Valve Compartment August 29, 2014

CA-M-20140912-7 6204 Ultrasonic Cleaner to Wash Down Pit September 12, 2014

CA-M-20140922-9 Pre-Decon of Cask Wash Down Pit September 22, 2014

CA-M-20141015-26

RB-200-I Survey inside bio- Routine October 15, 2014

CA-M-20141020-12 7404E Demin Valve Compartments October 20, 2014

CA-M-20141020-5 7404C Demin Valve Compartment October 20, 2014

CA-M-20141027-31

RB-200-I Survey inside bio- Routine October 27, 2014

Air Sample Results Sample Number Title Date

1410280415 EPHV8879A Valve Breach October 28, 2014

1410280855 Job Coverage - RB Seal Table October 28, 2014

1410281010 BB8948A - Lapping October 28, 2014

1410281720 BB8948A October 28, 2014

1410282358 Routine October 29, 2014

Miscellaneous Documents

Number Title Date

Callaway Radioactive Source Listing September 24, 2014

Sealed Source Leak Test Results January 7, 2014

Sealed Source Leak Test Results June 9, 2014

Miscellaneous Documents

Number Title Date Drawing

C252919 Reactor Building In-Core Instrumentation Tue Supports and Platforms

Section 2RS3: In-plant Airborne Radioactivity Control and Mitigation Procedures Number Title Revision

APA-ZZ-01004 General Instructions For Donning And Removing Respiratory Equipment 6

HDP-ZZ-08000

Respiratory Protection Program 22

HTP-ZZ-08300-DTI-Airlpak75 Scott Air Pak 75 SCBA Respirator Inspection and Storage 7

HTP-ZZ-08300-DTI-Skapak Skapak AT SCBA Respirator Storage and Inspection 4

HTP-ZZ-08501-DTI-Air Test Testing of Breathing Air 4

Callaway Action Requests

201004639

201305409

201308548

201403168

201403872

201406718

201406788

Lesson Plans and Training Guidance Number Title Date T210.0002 Air Pak Proficiency Demonstration March 2, 2011 T64.JT286 RP Just-In-Time Training - Delta Suit User

Breathing Air Test Results

Title Date

Breathing Air Sample Data Sheet September 18, 2014

Breathing Air Sample Data Sheet July 11, 2014

Breathing Air Sample Data Sheet April 7, 2014

Training Certificates Number Title Date Technician A Air-Pak 2.2/3.0/4.5/Fifty/75 SCBA Maintenance and Overhaul September 20, 2014 Technician B Air-Pak 2.2/3.0/4.5/Fifty/75 SCBA Maintenance and Overhaul September 20, 2014 Technician C Air-Pak 2.2/3.0/4.5/Fifty/75 SCBA Maintenance and Overhaul September 22, 2013

Miscellaneous

Title Date Delta Protection Mururoa Blu Single-Use Respiratory Protection Suit (TAC No

MD 3333) Approval December 18, 2006 Operations Crew Roster October 30, 2014 Respiratory Protection Equipment Inspection Records May 6, 2014 2014 On-Shift (Operations) Crew Schedule

Section 4OA1: Performance Indicator Verification Miscellaneous Title Date Consolidated Data Entry 4.0,

MSPI Derivation Report, MSPI High Pressure Injection System, Unavailability Index September 2014 Consolidated Data Entry 4.0, MSPI Derivation Report, MSPI High Pressure Injection System, Unreliability Index September 2014 Consolidated Data Entry 4.0, MSPI Derivation Report, MSPI Residual Heat Removal System, Unavailability Index September 2014 Consolidated Data Entry 4.0, MSPI Derivation Report, MSPI Residual Heat Removal System, Unreliability Index September 2014 NRC Performance Indicator Transmittal Report, Fourth Quarter 2013, Mitigating Systems Cornerstone January 9, 2014 NRC Performance Indicator Transmittal Report, First Quarter 2014, Mitigating Systems Cornerstone April 9, 2014 NRC Performance Indicator Transmittal Report, Second Quarter 2014, Mitigating Systems Cornerstone July 1, 2014 NRC Performance Indicator Transmittal Report, Third Quarter 2014, Mitigating Systems Cornerstone October 8, 2014

Section 4OA2: Identification and Resolution of Problems Procedures Number Title Revision/Date

APA-ZZ-00500 Corrective

Action Program 61

ODP-ZZ-00001, Addendum 12 Operator Burdens and Workarounds 4

ODP-ZZ-00002 Equipment Status Control 75

ODP-ZZ-00008 Night Orders - Standing Orders - Operations Information Reports 13

OSP-BB-VL006 RCS Pressure Isolation Valves Inservice Tests - IPTE 43

OSP-EP-00001, Checklist 1 Accumulator Isolation Valves Pwr Removed Verification - 17566 November 16, 2014

OTG-ZZ-00001, Addendum 2 Safety Injection Accumulator Preparation 3

Callaway Action Requests

201403798

201403895

201404092

201404344

201408530

Jobs

13507340

14509428

Miscellaneous

Title Date Adverse and Emerging Trends Between 2014/06/01 and 2014/12/02 December 2, 2014 Corrective Action Review Board Package December 17, 2014

Section 4OA3: Event Follow-Up Procedures Number Title Revision/ Date

OTG-ZZ-00008 Normal Unit Recovery Guideline Following Reactor Trip 19

APA-ZZ-00542, Attachment 2 Post Transient Evaluation Report - Trip Number 83 December 4, 2014

Callaway Action Requests

201408897

201408898

201408899

Miscellaneous

Title Date Sequence of Events Log December 3, 2014 DNA History Plot - Main Turbine Speed December 3, 2014 DNA History Plot - Steam Generator Narrow Range Water Level December 3, 2014 Various Photos of Damaged Cables December 4, 2014

Section 4OA5: Other Activities Procedures/Work Instructions Number Title Revision/Date

N-8080-10 Technical Manufacturing Program for Forged Part for Closure Head B

APA-ZZ-01105 Site Access 38

APA-ZZ-01109 Safety/Security Interface Program 2

SDP-CP-00001 Site Protective Strategy 17

SDP-PI-DOORS Posting Portals and Barriers 14

SDP-PI-VEHAC Vehicle Access/Control 36

SDP-SF-00011 Compensatory Measures 22 51-9217229-003 Rigging Plan - Callaway Reactor Vessel Closure Head Replacement January 30, 2014 51-9093520-002 Callaway Plant Unit 1 Heavy Rigging January 30, 2014 03-9199483 Callaway Site Requirements Document 2014 RVCH Replacement

RF-20 0 51-9217961-000 Cribbing Reference Information - Callaway Unit 1 Reactor Vessel Closure Head Replacement

January 30, 2014 2348-P4 Procedure to Downend Replacement Reactor Vessel Closure Head Outside Reactor Building February 27, 2014 15005435-W-M01 Delivery of 160T Replacement Reactor Vessel Closure Head April 17, 2014 2348-P7 Procedure to Install Bigge Hatch Runway System October 22, 2013

Procedures/Work Instructions Number Title Revision/Date 2348-P6 Procedure to Install Upending Frame Inside Reactor Building February 27, 2014 50-9217796-000 Callaway Unit 1 Pre-Outage Master Traveler May 21, 2014 2348-P10 Procedure to Move Replacement Reactor Vessel Closure Head into Reactor Building February 28, 2014 02348-D3 Multi-Axle Transporter Land "G" Forces Pre-Outage Transportation May 26, 2014 02348-D2 Multi-Axle Transporter Land "G" Forces for Onsite Transportation April 30, 2014 03-9216566 Callaway Unit 1 Placement and Storage of Old Reactor Vessel Closure Head in Old Steam Generator Storage Facility 0 2348-P12.1 Procedure for Pre-Outage Work Load Testing April 4, 2014 2348-P5 Procedure to Remove Missile Shield from Reactor Building April 4, 2014 2348-P8 Procedure to Remove Old Reactor Vessel Closure Head from Reactor Building February 27, 2014 51-9179592-001 Requirements for Temporary Assembly Building at Callaway Energy Center October 18, 2010 03-9209907 Callaway Unit 1 Rolling Missile Shield Disassembly and Removal Procedure 1 2348-P13 Procedure for Replacement Reactor Vessel Closure Head Barge Offload 1 01-9218057-001 Disassembly of the Tooling for Transporting the Reactor Vessel Head with Control Rod Drive Mechanisms Callaway 1 B 2348-P2 Procedure to Upend Replacement Reactor Vessel Closure Head Shipping Package and Place in Head Assembly Building May 7, 2014 2348-P11 Procedure to Transport Integrated Head Assembly Pieces into Reactor Building February 28, 2014 2348-P1 Procedure to Transport Replacement Reactor Vessel Closure Head Shipping Package from Barge Offload Site to Head Assembly Building 2 2348-P3 Procedure to Transport Replacement Reactor Vessel Closure Head to Reactor Building February 28, 2014

Procedures/Work Instructions Number Title Revision/Date 2348-P9 Procedure to Upend Old Reactor Vessel Closure Head and Transport to Old Steam Generator Storage Facility February 27, 2014

APA-ZZ-00365, Addendum L Callaway Lifting Operations 14

APA-ZZ-00365 Addendum R Callaway Rigging Operations 8

ETP-BB-03147 Addendum 1 Installation of Reactor Vessel Head Lifting Rig 5

ETP-BB-03147 Reactor Vessel Head Removal-IPTE 20

ETP-BB-03154 Reactor Vessel Head Installation-IPTE 19

MP-127-003 Areva Integrated Head Tripod Load Testing 0

EDP-ZZ-04060 Post-Design and Configuration Change Testing 6

MP-10-0005 Post-Change Test Plan - Reactor Pressure Boundary Leakage 0

ESP-SF-00001 Rod Drop Testing Using the Plant Computer - IPTE 24

ESP-SQ-00001 Loose Parts Background Noise 13

ESP-SQ-00002 Loose Parts Monitor Baseline Data Acquisition 2

ESP-ZZ-00020 Rod Position Indication Checkout - IPTE 15

ESP-ZZ-00032 Test Procedure for Automated Measurement of Drop Times of Control and Shutdown Rods in Westinghouse PWRS WITH DRPI Power-On 2

ETP-BB-03159 Reactor Vessel Head Electrical Interference Installation 15

ISL-SQ-00Y64 Loose Parts Monitor System Cabinet SQ064 Loop Calibration 24

ITM-SF-00002 DRPI Troubleshooting Guidelines - Cable and Detector Coil Continuity/Integrity Testing 6

OSP-BB-V002B Reactor Vessel Head Vent Valves Inservice Test 14

OSP-SF-00002 Control Rod Partial Movement 22

OSP-SH-00001 PAM Channel Check 28

QCP-ZZ-05041 Visual Examination to ASME

VT-2 26

ESP-ZZ-00020 Rod Position Indication Checkout - IPTE 15

OTS-SF-00002 Control Rod Exercising for CRDM Crud Mitigation Mode 5 1

Procedures/Work Instructions Number Title Revision/Date

OTS-SF-00003 Control Rod Exercising for CRDM Crud Mitigation Mode 3 0

RP-DTI-Job-Cov-Refuel-Mtc-Guide Refueling Maintenance Guide 8

Radiation Work Permits Number Title Revision

13004902 RP and Decon Support for RVCH Replacement 1

13004889 Low Dose Disassembly/Reassembly of RVCH 1 13004898500 Remove, Transport, Place ORVCH in Storage 0 13004895500 High Dose Activities for Demolition of ORVCH 0 13004895302 Scaffold for Reactor Head Replacement 2 13004889450 Temporary Shield Reactor Vessel Closure Head on Head Stand 0

13004895 Clean Area Work for Reactor Head Replacement 0

Survey Documentation Number Title Date

CA-M-2013-0503-14 RBHEAD - Reactor Head May 3, 2013

CA-M-2014-0401-2 Old Steam Generator Storage Facility April 1, 2014

CA-M-2014-1022-22 RB2047 Reactor Head Survey October 22, 2014

CA-M-2014-1023-10 Reactor Head Scaffold Levels 1 and 2 October 23, 2014

CA-M-2014-1024-11 RB2047 Reactor Head Scaffold Levels 3 and 4 October 24, 2014

CA-M-2014-1031-28 Old Steam Generator Storage Facility October 31, 2014

CA-M-2014-1102-15 Old Steam Generator Storage Facility Old Reactor Vessel Closure Head Vault Plant Side South November 1, 2014

Calculations

Number Title Revision/Date 32-9214732-000 Callaway Old Reactor Vessel Closure Head Storage Floor Loading Evaluation January 30, 2014 32-9212628-000 Callaway Reactor Vessel Closure Head Haul Route Evaluation January 30, 2014

Calculations

Number Title Revision/Date 32-9216352-000 Callaway Reactor Vessel Closure Head Outside Lift System Foundation January 30, 2014 32-9220596-000 Callaway Reactor Vessel Closure Head Reactor Building Floor Loading Justification January 30, 2014 02348-D1 Design Basis - Engineering, Rigging and Transport Services, Reactor Vessel Head Replacement Project - Callaway May 26, 2014 C7.2 Head Assembly Building Cribbing - Concrete Stools 0 C1.1 Replacement Reactor Vessel Closure Head Shipping Package Gantry Upending 3 C2.1 Hatch Transfer Runway 3 C3.2 Inside Upending Frame 4 C2.4 Missile Shield Cart 2 C4.4 Missile Shield Lift Rigging 1 C4.5 Mobile Crane Setup 2 C5.1 Old Reactor Vessel Closure Head & Replacement Reactor Vessel Closure Head Trailer Configuration 2 32-9216352-000 Callaway Reactor Vessel Closure Head Outside Lift System Foundation 18 C5.2 Old Reactor Vessel Closure Head & Replacement Reactor Vessel Closure Head Transporter Tiedowns 2 C4.1.1 Old Reactor Vessel Closure Head Upend/Downend Rigging Attachments 3 C3.1 Outside Upending Frame 1 C6.1 Pre-Outage Work Load Testing 0 C5.3 Replacement Reactor Vessel Closure Head Shipping Package Trailer Configuration 4 C5.4 Replacement Reactor Vessel Closure Head Shipping Package Transporter Tiedowns 3 C3.3 Replacement Reactor Vessel Closure Head Staging Location Cribbing 1 C4.1.2 Replacement Reactor Vessel Closure Head Upend/Downend Rigging Attachments 1 C2.2 Reactor Vessel Closure Head Cart 1 C4.3 Reactor Vessel Closure Head Horizontal Rigging 1

Calculations

Number Title Revision/Date C4.2 Reactor Vessel Closure Head Vertical & Upend Rigging 1 C2.3 Utility Cart 1 32-9180957-001 Callaway Reactor Vessel Head Drop Model 17 32-9183546-000 Callaway Reactor Vessel Head Drop Analysis 17

BB-18 Evaluate the Impact to the Existing Head Drop Analyses Due to Integrated Head Assembly Installation 0 51-9201488-001 Callaway Upper Integrated Head Assembly Safe Load Path Evaluation November 20, 2012 33-9132947-000 Callaway Plant Unit 1 Reactor Head Replacement - Integrated Head Assembly Design Report 0 51-9217961-000 Cribbing Reference Information - Callaway Unit 1 Reactor Vessel Closure Head Replacement January 30, 2014 24430-CALC-N-004 Callaway Reactor Vessel Head Storage Building Shielding Requirements 0

Drawings

Number Title Revision 1HB1.30S0 Closure Head Shipping General Arrangement 2 02-9194984B-000 Cribbing for Replacement Reactor Closure Head Staging 1 02-9194958B-000 Downend and Staging of Replacement Reactor Vessel Closure Head 1 9223154B Callaway Nuclear Plant Reactor Vessel Closure Head Replacement Reactor Building Equipment Hatch Access Platform Stiffener Plate Notes 0 9223156B Callaway Nuclear Plant Reactor Vessel Closure Head Replacement Reactor Building Equipment Hatch Access Platform Stiffener Plate Sections 0 9223155B Callaway Nuclear Plant Reactor Vessel Closure Head Replacement Reactor Building Equipment Hatch Access Platform Stiffener Plates Plan 0 02-9194973B-000 Hatch Transfer Runway Plan View 3 02-9194983B-001 Inside Containment Upend/Downend Frame Isometric View 2 02-9194961B-000 Install of Replacement Reactor Vessel Closure Head into Containment

1 02-9195006B-000 Lift Exclusion Zones

1

Drawings

Number Title Revision 02-9194977B-000 Missile Shield Cart Assembly General Arrangement - Isometric View 1 02-9217364D Callaway Unit 1 Rolling Missile Shield End Truck Restraint Tie Down Plates 0 02-9216633E Callaway Unit 1 Rolling Missile Shield End Truck Temporary Restraints 0 52 Missile Shield Lifting Rigging Orthographic Views 2 02-9195007B-000 Mobile Crane Mat Layout Bottom Level Plan 1 02-9194962B-000 Move Integrated Head Components into Containment Equipment Lift onto Access Platform 1 02-9194980B-000 Reactor Vessel Head Replacement Isometric View Old Reactor Vessel Head Upend/Downend System 1 02-9194982B-000 Outside Upending Frame Isometric View 1 02-9195000B-001 Pre-Outage Work Equipment Testing Gantry Link/HGL-2 1 02-9194963B-000 Removal of Missile Shield for Reactor Building Plan View - Cart Staging 1 02-9194959B-000 Removal of Old Reactor Vessel Closure Head from Containment Vertical Lift Rigging Install 1 81 Replacement Reactor Vessel Closure Head Downend Frame Test Assembly Isometric Views 0 02-9194996B-001 General Arrangement - Transportation Replacement Reactor Vessel Closure Head Shipping Package Barge Offload 2 02-9194957B-001 Replacement Reactor Vessel Closure Head Shipping Package Upending and Storage Upending general Arrangement 2 02-9194995B-000 Replacement Reactor Vessel Closure Head Shipping Package Transport to Head Assembly Building Haul Route General Arrangement 4 02-9194981B-001 Reactor Vessel Head Replacement Isometric View Replacement Reactor Vessel Closure Head Upend/Downend System 1 02-9194967B-001 Replacement Reactor Vessel Closure Head Shipping Package Gantry Assembly - Plan 3 02-9194974B-000 Reactor Vessel Closure Head Runway Cart Assembly General Arrangement - Isometric View 2

Drawings

Number Title Revision 02-9194988B-000 Reactor Vessel Closure Head Horizontal Rigging - Inside Isometric View 1 02-9194989B-000 Reactor Vessel Closure Head Vertical & Upend/Downend Rigging - Inside Isometric View 1 02-9194994B-001 Reactor Vessel Closure Head Onsite Transport Heavy Haul Route 4 02-9194997B-000 Reactor Vessel Closure Head Horizontal Rigging - Outside Isometric View 1 02-9194998B-000 Reactor Vessel Closure Head Vertical and Upend/Downend Rigging - Outside Isometric View 1 02-9194975B-000 Utility Cart Assembly General Arrangement - Isometric View 1 02-9194960B-000 Upend Old Reactor Vessel Closure Head and Load to Transport Plan View 1 8600-X-88100 Property-Site Layout Owner Controlled Area and Surrounding Area 66 8600-X-90706 Old Steam Generator Storage Facility Concrete Plans 0 8600-X-90707 Old Steam Generator Storage Facility Concrete and Reinforcing Sections 0 1HB1.1001 Closure Head Cladding Machining and Weld Lugs 6

1HB1.1001 Closure Head Cladding Machining and Weld Lug 7 BUMPCA/NCC0530 CETNA Assembly

BUMPCA/NCC0430 CRDM Assembly and Welding

Callaway Action Requests

200708758

201002759

201003045

201104123

201104568

201104569

201204894

201204623

201204661

201204801

201205101

201205138

201205302

201305503

201305327

201305423

201305426

201305556

201305604

201305676

201305731

201305742

201305768

201305807

201305823

201305856

201315676

201406137

201406201

201406264

201406264

201407696

201408368

201408381

201408385

Corrective Action Requests

NCR-1HB1-007

NCR-1HB-011

NCR-1HB-012

EHM-203-10

ALARA Plans RWP Number Title Date

13004889 ALARA Plan - Level 1 June 17, 2014

13004895 ALARA Plan - Level 1 June 17, 2014 13004889500 ALARA Planning Open Item Checklist May 29, 2014 13004889500 ALARA Plan - Level 2 May 29, 2014 13004895302 ALARA Planning Open Item Checklist May 28, 2014 13004895302 ALARA Plan - Level 2 June 16, 2014 13004898500 ALARA Planning Open Item Checklist June 16, 2014 13004898500 ALARA Plan - Level 2 June 16, 2014

ALARA In Progress Reviews 2014-071 Work In-Progress ALARA Review - R20-55220 New Reactor Vessel Closure Head

RF 20 Head Replacement August 20, 2014 2014-071 Work In-Progress ALARA Review - R20-55220 New Reactor Vessel Closure Head

RF 20 Head Replacement October 26, 2014

Miscellaneous

Number Title Revision/Date 1CA2510 Applicability Determination - "Implement an Integrated Head Assembly in RF20 in Support of Replacement Reactor Vessel Head Project" 4 1CA2511 50.59 Screen - "Implement an Integrated Head Assembly in RF20 in Support of Replacement Reactor Vessel Head Project" 4 5CA2510 Applicability Determination - Replace Pressure Boundary Components Associated with Replacement Reactor Vessel Head 4 5CA2511 50.59 Screen - Replace Pressure Boundary Components Associated with Replacement Reactor Vessel Head 4 13-003 50.59 Evaluation - Implement an Integrated Head Assembly in RF20 in Support of Replacement Reactor Vessel Head Project 0

Miscellaneous

Number Title Revision/Date STARS - ENG - 5001 - 8.1 Engineering Disposition - Implement an Integrated Head Assembly in RF20 in support of Replacement Reactor Vessel Closure Head Project September 4, 2012 51-9185539-002 Callaway Unit 1 Reactor Vessel Closure Head ASME Code Reconciliation October 21, 2013

Contract Between Union Electric Company d/b/a Ameren UE and Areva NP Inc. for Reactor Vessel Head Replacement Services for the Callaway Nuclear Plant July 20, 2009 M-2012 (Q) Technical Specification for the Replacement Reactor Vessel Closure Head Project 5 33-9180485-002 ASME Design Report for Callaway Unit 1 Replacement Reactor Vessel Closure Head November 20, 2012 87-9182060-001 Contract Variation Approval Request August 13, 2012 87-9200533-000 Contract Variation Approval Request March 14, 2013

TEC-13-076 Final Report of the Operational Inspection Performed on the Replacement Reactor Vessel Closure Head for Callaway Unit 1 2

TEC-13-114 Final Report of Ultrasonic Inspection, Eddy Current Inspection, and Visual Inspection on the Replacement Reactor Vessel Closure Head for Callaway 1 Nuclear Power Plant 2 N-8080-10 Technical Manufacturing Program for Forged Part for Closure Head B

JQA-10-129 Certified Material Test Report June 29, 2013

QP-08 Closure Head Hydrostatic Test and Pre-service Inspection 2 BUQRCA/NCC002 End of Manufacturing Report CC/CA001 for Thermocouples for Callaway Unit 1 A 1HB1/015 Welding Material Receiving Inspection Report 02 1HB1/016 Welding Material Receiving Inspection Report 2 1HB1/017 Welding Material Receiving Inspection Report 1 1HB1/018 Welding Material Receiving Inspection Report 2 1HB1/020 Welding Material Receiving Inspection Report 0 1HB1/021 Welding Material Receiving Inspection Report 0 1HB1/022 Welding Material Receiving Inspection Report 0

Miscellaneous

Number Title Revision/Date

PQR-1128 Procedure Qualification Record

January 3, 2005 1HB1

WT 207 Welding Procedure Specification 4

QP-06 Shrink Fitting and Head Penetration Nozzle Weld (ILH, Thermal, Vent, and Reactor Vessel Level Indicating System) Quality Assurance Data Package 4 1HB1CS401 Liquid Penetrant Examinations - Replacement Reactor Vessel Closure Head for Callaway Unit 1 3

PQR 1068 Procedure Qualification Record October 10, 2002 1HB1

WT 109 Welding Procedure Specification 2

QP-06 SRP02 Standard Repair Process Cycle for the Head Penetration Nozzles Quality Assurance Data Package 1

QP-06 SRP03 Repair Process to J-Groove 1B-10.08.XX Quality Assurance Data Package 1

QP-06 SRP03 Repair Process to ILH No. 66 Quality Assurance Data Package 0

PQR-966 Welding Procedure Qualification between Inconel Buttering on P3 Gr3 with Shielded Metal Arc Welding Process and Inconel 690 material (GTAW-SMAW Process) April 30, 1999 23-9215950-001 Quality Assurance Data Package for Control Rod Drive Mechanism Type L106-A1 Two Rod Travel Housings (Spare Parts) Callaway Plant Unit 1 April 23, 2014 23-9212915-001 Quality Assurance Data Package for Control Rod Drive Mechanism Assembly Type L106-A1 Callaway Plant Unit 1 April 23, 2014

FN9-2078 Certified Material Test Report - Replacement Reactor Vessel Closure Head C 1HB1CS401 Liquid Penetrant Examinations - Replacement Reactor Vessel Closure Head for Callaway Unit 1 4 1HB1FS502 Painting, Final Cleaning and Preparation for Transport - Replacement Reactor Vessel Closure Head for Callaway Unit 1 3 02-9195006B-000 Lift Exclusion Zones, Exterior Exclusion Zones, Reactor Vessel Head Replacement, Callaway Power Plant, Unit 1/AREVA NP, Inc. April 8, 2014

RRVCH-ANP 13-0418 Ameren Response to AREVA 13-02784, Request for Design Input to Radiological Drop Evaluation for the Old Reactor Vessel Closure Head November 15, 2013

Miscellaneous

Number Title Revision/Date T64.03040.8 Radiation Protection Continuing Training:

Reactor Head Replacement July 18, 2013 T64.03040.8 Radiation Chemistry Retraining - RP Operations:

Reactor Head Modification Work June 18, 2014

Attachment 2

The following items are requested for the Occupational Radiation Safety Inspection at Callaway Plant October 27-31, 2014 Integrated Report

2014005

Inspection areas are listed in the attachments below.

Please provide the requested information on or before September 25, 2014.

Please submit this information using the same lettering system as below.

For example, all contacts and phone numbers for Inspection Procedure 71124.01 should be in a file/folder titled "1- A," applicable organization charts in file/folder "1- B," etc.

If information is placed on ims.certrec.com, please ensure the inspection exit date entered is at least 30 days later than the onsite inspection dates, so the inspectors will have access to the information while writing the report.

In addition to the corrective action document lists provided for each inspection procedure listed below, please provide updated lists of corrective action documents at the entrance meeting.

The dates for these lists should range from the end dates of the original lists to the day of the entrance meeting.

If more than one inspection procedure is to be conducted and the information requests appear to be redundant, there is no need to provide duplicate copies.

Enter a note explaining in which file the information can be found.

If you have any questions or comments, please contact Larry Ricketson at (817) 200-1165 or Larry.Ricketson@nrc.gov.

PAPERWORK REDUCTION ACT STATEMENT This letter does not contain new or amended information collection requirements subject to the Paperwork Reduction Act of 1995 (44 U.S.C. 3501 et seq.). Existing information collection requirements were approved by the Office of Management and Budget, control number 3150-0011.

1. Radiological Hazard Assessment and Exposure Controls (71124.01) and Performance Indicator Verification (71151) Date of Last Inspection: April 15, 2013

A. List of contacts and telephone numbers for the Radiation Protection Organization staff and technicians B. Applicable organization charts C. Audits, self-assessments, and LERs written since date of last inspection, related to this inspection area D. Procedure indexes for the radiation protection procedures E. Please provide specific procedures related to the following areas noted below.

Additional Specific Procedures may be requested by number after the inspector reviews the procedure indexes.

1. Radiation Protection Program Description 2. Radiation Protection Conduct of Operations 3. Personnel Dosimetry Program 4. Posting of Radiological Areas 5. High Radiation Area Controls 6. RCA Access Controls and Radworker Instructions 7. Conduct of Radiological Surveys 8. Radioactive Source Inventory and Control 9. Declared Pregnant Worker Program F. List of corrective action documents (including corporate and subtiered systems) since date of last inspection a. Initiated by the radiation protection organization b. Assigned to the radiation protection organization

NOTE:

The lists should indicate the significance level of each issue and the search criteria used.

Please provide in document formats which are "searchable" so that the inspector can perform word searches. If not covered above, a summary of corrective action documents since date of last inspection involving unmonitored releases, unplanned releases, or releases in which any dose limit or administrative dose limit was exceeded (for Public Radiation Safety Performance Indicator verification in accordance with

IP 71151) G. List of radiologically significant work activities scheduled to be conducted during the inspection period (If the inspection is scheduled during an outage, please also include a list of work activities greater than 1 rem, scheduled during the outage with the dose estimate for the work activity.) H. List of active radiation work permits I. Radioactive source inventory list a.

All radioactive sources that are required to be leak tested b.

All radioactive sources that meet the 10 CFR Part 20, Appendix E, Category 2, and above threshold.

Please indicate the radioisotope, initial and current activity (w/assay date), and storage location for each applicable source.

J.

The last two leak test results for the radioactive sources inventoried and required to be leak tested (If applicable, specifically provide a list of all radioactive source(s) that have failed its leak test within the last two years.)

K. A current listing of any non-fuel items stored within your pools, and if available, their appropriate dose rates (Contact / @ 30cm)

L. Computer printout of radiological controlled area entries greater than 100 millirems since the previous inspection to the current inspection entrance date (The printout should include the date of entry, some form of worker identification, the radiation work permit used by the worker, dose accrued by the worker, and the electronic dosimeter dose alarm setpoint used during the entry (for Occupational Radiation Safety Performance Indicator verification in accordance with

IP 71151)).

3.

In-Plant Airborne Radioactivity Control and Mitigation (71124.03)

Date of Last Inspection: August 27, 2012

A. List of contacts and telephone numbers for the following areas: 1. Respiratory Protection Program 2. Self-contained breathing apparatus

B. Applicable organization charts C. Copies of audits, self-assessments, vendor or NUPIC audits for contractor support (SCBA), and LERs, written since date of last inspection related to:

1. Installed air filtration systems 2. Self-contained breathing apparatuses

D. Procedure index for: 1. Use and operation of continuous air monitors 2. Use and operation of temporary air filtration units

3. Respiratory protection E. Please provide specific procedures related to the following areas noted below.

Additional specific procedures may be requested by number after the inspector reviews the procedure indexes.

1. Respiratory protection program 2. Use of self-contained breathing apparatuses

3. Air quality testing for SCBAs

4. Use of installed plant systems, such as containment purge, spent fuel pool ventilation, and auxiliary building ventilation F. A summary list of corrective action documents (including corporate and subtiered systems) written since date of last inspection, related to the Airborne Monitoring program including: 1. Continuous air monitors 2. Self-contained breathing apparatuses

3. Respiratory protection program NOTE:

The lists should indicate the significance level of each issue and the search criteria used.

Please provide in document formats which are "searchable" so that the inspector can perform word searches. G. List of SCBA qualified personnel - reactor operators and emergency response personnel

H. Inspection records for self-contained breathing apparatuses (SCBAs) staged in the plant for use since date of last inspection I. SCBA training and qualification records for control room operators, shift supervisors, STAs, and OSC personnel for the last year

A selection of personnel may be asked to demonstrate proficiency in donning, doffing, and performance of functionality check for respiratory devices

J. List of respirators (available for use) by type (APR, SCBA, PAPR, etc.), manufacturer, and model Attachment 3

PAPERWORK REDUCTION ACT STATEMENT

This letter does not contain new or amended information collection requirements subject to the Paperwork Reduction Act of 1995 (44 U.S.C. 3501 et seq.).

Existing information collection requirements were approved by the Office of Management and Budget, Control Number 3150-0011.

The NRC may not conduct or sponsor, and a person is not required to respond to, a request for information or an information collection requirement unless the requesting document displays a currently valid Office of Management and Budget control number.

In accordance with 10 CFR 2.390 of the NRC's "Rules of Practice," a copy of this letter and its enclosure will be available electronically for public inspection in the NRC Public Document Room or from the Publicly Available Records (PARS) component of NRC's document system (ADAMS).

ADAMS is accessible from the NRC web site at http://www.nrc.gov/reading-rm/adams.html (the Public Electronic Reading Room). .

Information Request January 22, 2014 Notification of Inspection and Request for Information Callaway Plant NRC Inspection Reports 05000483/2014004 and 05000483/2014005

On October 20, 2014, reactor inspectors from the Nuclear Regulatory Commission's (NRC) Region IV office will perform the baseline inservice inspection at Callaway Plant, using NRC Inspection Procedure 71111.08, "Inservice Inspection Activities."

Experience has shown that this inspection is a resource intensive inspection both for the NRC inspectors and your staff.

In order to minimize the impact to your onsite resources and to ensure a productive inspection, we have enclosed a request for documents needed for this inspection.

These documents have been divided into two groups.

The first group (Section A of the enclosure) identified information to be provided prior to the inspection to ensure that the inspectors are adequately prepared.

The second group (Section B of the enclosure) identifies the information the inspectors will need upon arrival at the site.

It is important that all of these documents are up to date and complete in order to minimize the number of additional documents requested during the preparation and/or the onsite portions of the inspection.

We have discussed the schedule for these inspection activities with your staff and understand that our regulatory contact for this inspection will be Mr. Steve Petzel of your licensing organization.

The tentative inspection schedule is as follows:

Preparation week: October 13, 2014

Onsite weeks: October 20 - 31, 2014

Our inspection dates are subject to change based on your updated schedule of outage activities.

If there are any questions about this inspection or the material requested, please contact the lead inspector Wayne Sifre at (817) 200-1193 (Wayne.Sifre@nrc.gov).

A.1 ISI/Welding Programs and Schedule Information a) A detailed schedule (including preliminary dates) of: i. Non-destructive examinations planned for ASME Code Class Components performed as part of your ASME Section XI, risk informed (if applicable), and augmented inservice inspection programs during the upcoming outage. ii. Examinations planned for Alloy 82/182/600 components that are not included in the Section XI scope (If applicable) iii. Examinations planned as part of your boric acid corrosion control program

(Mode 3 walkdowns, bolted connection walkdowns, etc.) iv. Welding activities that are scheduled to be completed during the upcoming outage (ASME Class 1, 2, or 3 structures, systems, or components) b) A copy of ASME Section XI Code Relief Requests and associated NRC safety evaluations applicable to the examinations identified above. i.

A list of ASME Code Cases currently being used to include the system and/or component the Code Case is being applied to. c) A list of non-destructive examination reports which have identified recordable or rejectable indications on any ASME Code Class components since the beginning of the last refueling outage. This should include the previousSection XI pressure test(s) conducted during start up and any evaluations associated with the results of the pressure tests. d) A list including a brief description (e.g., system, code class, weld category, non-destructive examination performed) associated with the repair/replacement activities of any ASME Code Class component since the beginning of the last outage and/or planned this refueling outage. e) If reactor vessel weld examinations required by the ASME Code are scheduled to occur during the upcoming outage, provide a detailed description of the welds to be examined and the extent of the planned examination.

Please also provide reference numbers for applicable procedures that will be used to conduct these examinations. f) Copy of any 10 CFR Part 21 reports applicable to structures, systems, or components within the scope of Section XI of the ASME Code that have been identified since the beginning of the last refueling outage. g) A list of any temporary noncode repairs in service (e.g., pinhole leaks). h) Please provide copies of the most recent self-assessments for the inservice inspection, welding, and Alloy 600 programs

A.2 Reactor Pressure Vessel Head Documents requested under Inspection Procedure 71007, "Reactor Vessel Head Replacement Inspection."

A.3 Boric Acid Corrosion Control Program a) Copy of the procedures that govern the scope, equipment and implementation of the inspections required to identify boric acid leakage and the procedures for boric acid leakage/corrosion evaluation. b) Please provide a list of leaks (including code class of the components) that have been identified since the last refueling outage and associated corrective action documentation.

If during the last cycle, the unit was shutdown, please provide documentation of containment walkdown inspections performed as part of the boric acid corrosion control program.

A.4 Steam Generator Tube Inspections a) A detailed schedule of: i. Steam generator tube inspection, data analyses, and repair activities for the upcoming outage (if occurring). ii. Steam generator secondary side inspection activities for the upcoming outage (if occurring). b) Copy of SG history documentation given to vendors performing eddy current (ET) testing of the SGs during the upcoming outage. c) Copy of procedure containing screening criteria used for selecting tubes for in-situ pressure testing and the procedure to be used for in-situ pressure testing. d) Copy of previous outage SG tube operational assessment completed following ET of the SGs.

Also include a copy of the following documents as they become available: i. Degradation assessment ii. Condition monitoring assessment e) Copy of the document defining the planned SG ET scope (e.g., 100 percent of unrepaired tubes with bobbin probe and 20 percent sample of hot leg expansion transition regions with rotating probe) and identify the scope expansion criteria, which will be applied.

Also identify and describe any deviations in this scope or expansion criteria from the EPRI Guidelines. f) Copy of the document describing the ET acquisition equipment to be applied including ET probe types.

Also identify the extent of planned tube examination coverage with each probe type (e.g. rotating probe -0.080 inches, 0.115 inches pancake coils and mid-range +point coil applied at the top-of-tube-sheet plus 3 inches to minus 12 inches).

g) Identify and quantify any SG tube leakage experienced during the previous operating cycle.

Also provide documentation identifying which SG was leaking and corrective actions completed and planned for this condition. h) Copy of steam generator eddy current data analyst guidelines and site validated eddy current technique specification sheets.

Additionally, please provide a copy of EPRI Appendix H, "Examination Technique Specification Sheets," qualification records. i) Provide past history of the condition and issues pertaining to the secondary side of the steam generators (including items such as loose parts, fouling, top of tube sheet condition, crud removal amounts, etc.). j) Indicate where the primary, secondary, and resolution analyses are scheduled to take place.

A.5 Additional Information Related to all Inservice Inspection Activities a) A list with a brief description of inservice inspection, boric acid corrosion control program, and steam generator tube inspection related issues (e.g., condition reports) entered into your corrective action program since the beginning of the last refueling outage.

For example, a list based upon data base searches using key words related to piping or steam generator tube degradation such as: inservice inspection, ASME

Code,Section XI, non-destructive examination, cracks, wear, thinning, leakage, rust, corrosion, boric acid, or errors in piping/steam generator tube examinations. b) Provide training (e.g. Scaffolding, Fall Protection, FME, Confined Space) if they are required for the activities described in A.1 through A.4. c) Please provide names and phone numbers for the following program leads:

Inservice inspection (examination, planning) Containment exams Reactor pressure vessel head exams Snubbers and supports Repair and replacement program

Licensing

Site welding engineer Boric acid corrosion control program Steam generator inspection activities (site lead and vendor contact)

B. Information to be Provided Onsite to the Inspector(s) at the Entrance Meeting (October 20, 2014): B.1 Inservice Inspection / Welding Programs and Schedule Information a) Updated schedules for inservice inspection/non-destructive examination activities, including planned welding activities, and schedule showing contingency repair plans, if available.

b) For ASME Code Class welds selected by the inspector from the lists provided from section A of this enclosure, please provide copies of the following documentation for each subject weld: i. Weld data sheet (traveler). ii. Weld configuration and system location. iii. Applicable Code Edition and Addenda for weldment. iv. Applicable Code Edition and Addenda for welding procedures. v. Applicable welding procedures used to fabricate the welds. vi. Copies of procedure qualification records (PQRs) supporting the weld procedures from B.1.b.v. vii. Copies of welder's performance qualification records (WPQ). viii. Copies of the nonconformance reports for the selected welds (If applicable). ix. Radiographs of the selected welds and access to equipment to allow viewing radiographs (if radiographic testing was performed). x. Copies of the preservice examination records for the selected welds. xi. Readily accessible copies of non-destructive examination personnel qualifications records for reviewing. c) For the inservice inspection related corrective action issues selected by the inspectors from section A of this enclosure, provide a copy of the corrective actions and supporting documentation. d) For the non-destructive examination reports with relevant conditions on ASME Code Class components selected by the inspectors from Section A above, provide a copy of the examination records, examiner qualification records, and associated corrective action documents. e) A copy of (or ready access to) most current revision of the inservice inspection program manual and plan for the current interval. f) For the non-destructive examinations selected by the inspectors from section A of this enclosure, provide a copy of the non-destructive examination procedures used to perform the examinations (including calibration and flaw characterization/sizing procedures).

For ultrasonic examination procedures qualified in accordance with ASME Code,Section XI, Appendix VIII, provide documentation supporting the procedure qualification (e.g. the EPRI performance demonstration qualification summary sheets).

Also, include qualification documentation of the specific equipment to be used (e.g., ultrasonic unit, cables, and transducers including serial numbers) and non-destructive examination personnel qualification records.

B.2 Reactor Pressure Vessel Head (RPVH)

Documents

requested under Inspection Procedure 71007, "Reactor Vessel Head Replacement Inspection."

B.3 Boric Acid Corrosion Control Program a) Please provide boric acid walk down inspection results, an updated list of boric acid leaks identified so far this outage, associated corrective action documentation, and overall status of planned boric acid inspections. b) Please provide any engineering evaluations completed for boric acid leaks identified since the end of the last refueling outage.

Please include a status of corrective actions to repair and/or clean these boric acid leaks.

Please identify specifically which known leaks, if any, have remained in service or will remain in service as active leaks.

B.4 Steam Generator Tube Inspections a) Copies of the Examination Technique Specification Sheets and associated justification for any revisions. b) Please provide a copy of the eddy current testing procedures used to perform the steam generator tube inspections (specifically calibration and flaw characterization/sizing procedures, etc.). c) Copy of the guidance to be followed if a loose part or foreign material is identified in the steam generators. d) Identify the types of SG tube repair processes which will be implemented for defective SG tubes (including any NRC reviews/evaluations/approvals of this repair process).

Provide the flaw depth sizing criteria to be applied for ET indications identified in the SG tubes. e) Copy of documents describing actions to be taken if a new SG tube degradation mechanism is identified. f) Provide procedures with guidance/instructions for identifying (e.g. physically locating the tubes that require plugging) and plugging SG tubes. g) List of corrective action documents generated by the vendor and/or site with respect to steam generator inspection activities.

B.5 Codes and Standards a) Ready access to (i.e., copies provided to the inspector(s) for use during the inspection at the onsite inspection location, or room number and location where available): i. Applicable Editions of the ASME Code (Sections V, IX, and XI) for the inservice inspection program and the repair/replacement program.

ii. EPRI and industry standards referenced in the procedures used to perform the steam generator tube eddy current examination. b) Copy of the performance demonstration initiative (PDI) generic procedures with the latest applicable revisions that support site qualified ultrasonic examinations of piping welds and components (e.g.,

PDI-UT-1,

PDI-UT-2,

PDI-UT-3,

PDI-UT-10, etc.). c) EPRI and industry standard references in the site procedures used to perform the SG

tube eddy current examination, which includes EPRI documents:

TR-107621-R1, "Steam Generator Integrity Assessment Guidelines,"

TR-107620-R1, "Steam Generator In-Situ Pressure Test Guidelines," Steam Generator Management Program: Steam Generator Integrity

Assessment Guidelines, Part 10, and

1003138, "Pressurized Water Reactor Steam Generator Examination Guidelines." d) Boric Acid Corrosion Guidebook Revision 1 - EPRI Technical Report 1000975.