IR 05000321/2005002

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IR 05000321-05-002, IR 05000366-05-002 & IR 0720000360-05-001, on 01/01/2005 - 03/31/2005, Edwin I. Hatch Nuclear Plant Unit 1 and Unit 2, Radioactive Material Processing and Transportation
ML051190731
Person / Time
Site: Hatch, 07200036  Southern Nuclear icon.png
Issue date: 04/29/2005
From: Widmann M
NRC/RGN-II/DRP/RPB2
To: Sumner H
Southern Nuclear Operating Co
References
IR-05-001, IR-05-002
Download: ML051190731 (28)
v • d • e


Text

ril 29, 2005

SUBJECT:

EDWIN I. HATCH NUCLEAR PLANT - NRC INTEGRATED INSPECTION REPORT 05000321/2005002, 05000366/2005002, AND 072000036/2005001

Dear Mr. Sumner:

On March 31, 2005, the U. S. Nuclear Regulatory Commission (NRC) completed an inspection at your Edwin I. Hatch Nuclear Plant, Units 1 and 2. The enclosed integrated inspection report documents the inspection findings, which were discussed on April 7, 2005, with Mr. Dennis Madison and other members of your staff.

The inspection examined activities conducted under your license as they relate to safety and compliance with the Commissions rules and regulations and with the conditions of your license.

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

This report documents one NRC-identified finding of very low safety significance (Green) which was determined to involve a violation of NRC requirements. Because this violation is of very low safety significance and was entered into your corrective action program, the NRC is treating this violation as a non-cited violation (NCV) consistent with Section VI.A of the NRC Enforcement Policy. If you contest this NCV, you should provide a response within 30 days of the date of this inspection report, with the basis for your denial, to the United States Nuclear Regulatory Commission, ATTN: Document Control Desk, Washington DC 20555-0001; with copies to the Regional Administrator, Region II; the Director, Office of Enforcement, United States Nuclear Regulatory Commission, Washington, DC 20555-0001; and the NRC Resident Inspector at the Hatch Nuclear Plant.

In accordance with 10 CFR 2.390 of the NRCs Rules of Practice, a copy of this letter, its enclosures, and your response (if any) will be available electronically for public inspection in the

SNC 2 NRC Public Document Room or from the Publicly Available Records (PARS) component of the NRCs 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).

Sincerely,

/RA/

Malcolm T. Widmann, Chief Reactor Projects Branch 2 Division of Reactor Projects Docket Nos. 50-321, 50-366, 72-36 License Nos. DPR-57 and NPF-5

Enclosure:

Inspection Report 05000321/2005002, 05000366/2005002, and 072000036/2005001 w/Attachment: Supplemental Information

REGION II==

Docket Nos: 50-321, 50-366, 72-36 License Nos: DPR-57, NPF-5 Report No: 05000321/2005002, 05000366/2005002, and 072000036/2005001 Licensee: Southern Nuclear Operating Company, Inc. (SNC)

Facility: Edwin I. Hatch Nuclear Plant Location: P.O. Box 2010 Baxley, Georgia 31515 Dates: January 1 - March 31, 2005 Inspectors: D. Simpkins, Senior (Sr.) Resident Inspector J. Hickey, Resident Inspector A. Vargas-Mendez, Reactor Inspector (Section 1R08)

S. Rose, Sr. Operations Engineer (Section 1R11)

G. Kuzo, Sr. Health Physicist (Sections 2OS1, 2PS2, and 4OA1)

A. Nielsen, Health Physicist (Sections 2OS2 and 4OA5)

Approved By: Malcolm T. Widmann, Chief Reactor Projects Branch 2 Division of Reactor Projects Enclosure

SUMMARY OF FINDINGS

IR 05000321/2005002, 05000366/2005002, 72000036/2005001; 01/01/2005 - 03/31/2005,

Edwin I. Hatch Nuclear Plant Unit 1 and Unit 2, Radioactive Material Processing and Transportation The report covered a three-month period of inspection by resident inspectors and regional health physicists, a reactor inspector, and an operations engineer. One Green non-cited violation was identified. The significance of most findings is indicated by their color (Green,

White, Yellow, or Red) using Inspection Manual Chapter (IMC) 0609, Significance Determination Process (SDP). Findings for which the SDP does not apply may be Green or be assigned a severity level after management review. The NRC's program for overseeing the safe operation of commercial nuclear power reactors is described in NUREG-1649, Reactor Oversight Process, Revision 3, dated July, 2000.

NRC-Identified and Self-Revealing Findings

Cornerstone: Public Radiation Safety

Green.

An NRC-identified non-cited violation of 10 CFR 71.5 was identified for failure to implement current package design specifications for proper closing of Type A shipping packages (control rod drive mechanism (CRDM) shipment boxes) as required by DOT regulations. Specifically, for Type A packages containing CRDM equipment shipped between January 2003 and February 2005, the licensee failed to prepare the package closures in accordance with vendor package specifications as required by 49 CFR 173.475(e).

This finding is more than minor because it is associated with the public radiation cornerstone program and process attribute and it affected the cornerstone objective to ensure adequate protection of public health and safety from exposure to radioactive material released into the public domain. The issue was reviewed using the Public Radiation Safety Significance Determination Process and was determined to be of very low safety significance (Green) because a radiation limit was not exceeded nor was the packaging breached. In addition, previous shipments made by the licensee had arrived at their destination with no identified degradation of the subject packaging and immediate corrective actions assured that on-going CRDM equipment packages were prepared properly prior to shipment. (Section 2PS2)

Licensee-Identified Violations

None.

REPORT DETAILS

Summary of Plant Status

Unit 1 operated at or near full Rated Thermal Power (RTP) during this inspection period, with the exception of a scheduled outage from January 3 to January 10 to repair drywell cooler leakage.

Unit 2 operated at or near full RTP until February 5 when the unit was shutdown for a scheduled refueling outage. The unit was restarted on March 12 and achieved RTP on March

REACTOR SAFETY

Cornerstones: Initiating Events, Mitigating Systems, and Barrier Integrity

1R04 Equipment Alignment

a. Inspection Scope

Partial System Walkdowns. The inspectors performed partial walk-downs of the following four systems when the redundant trains were removed from service. The inspectors checked system valve positions, electrical breaker positions, and operating switch positions to evaluate the operability of the redundant trains or components by comparing the position listed in the system operating procedure to the actual position.

Documents reviewed are listed in the Attachment.

  • 2C Emergency Diesel Generator (EDG) during a 2A EDG Surveillance Complete System Walkdown. The inspectors performed a complete walkdown of the following system. The inspectors performed a detailed check of valve positions, electrical breaker positions, and operating switch positions to evaluate the operability of the redundant trains or components by comparing the required position in the system operating procedure to the actual position. The inspectors also interviewed personnel and reviewed control room logs to verify that alignment and equipment discrepancies were being identified and appropriately resolved. Documents reviewed are listed in the

.

  • Decay Heat Removal/Unit-1 Fuel Pool Cooling

b. Findings

No findings of significance were identified.

1R05 Fire Protection

a. Inspection Scope

Fire Area Tours. The inspectors toured 12 risk significant areas to assess the material condition of the fire protection and detection equipment and to verify fire protection equipment was not obstructed. The inspectors reviewed procedure 40AC-ENG-008-OS, Fire Protection Program, and conducted area walk-downs to assess the licensee's control of transient combustibles. The inspectors also reviewed the Site Fire Hazards Analysis and applicable Pre-fire Plan drawings to verify that the necessary fire fighting equipment, such as fire extinguishers, hose stations, ladders, and communications equipment, were in place. Documents reviewed are listed in the Attachment.

  • Unit 1 Southeast RHR and CS Room
  • Unit 1 RCIC Pump and Turbine Room
  • Unit 1 NE RHR and CS Room
  • Unit 1 Working Floor and Air Supply Room 185'
  • Unit 1 CRD Area
  • Unit 2 Recirculation Motor Generator (MG) Set Rooms
  • Unit 1 Working Floor and MG Set Rooms 158'
  • Unit 1 Standby Gas and Air Conditioning 164'
  • Diesel Fuel Oil Storage Tanks

b. Findings

No findings of significance were identified.

1R07 Heat Sink Performance

a. Inspection Scope

Annual Resident Review. The inspectors reviewed the results of the licensee inspection of the 2B RHR heat exchanger. The inspectors verified implementation of procedure 42IT-TET-012-2S, Plant Service Water and RHR Service Water Piping Inspection Procedure. Documents reviewed are listed in the Attachment.

b. Findings

No findings of significance were identified.

1R08 Inservice Inspection (ISI) Activities

a. Inspection Scope

The inspectors observed the following Unit 2 ISI activities, reviewed ISI procedures, and reviewed selected ISI records, associated with risk significant structures, systems, and components. The observations and records were compared to the requirements specified in the Technical Specifications (TS) and the ASME Boiler and Pressure Vessel Code, 1989 Edition, to verify compliance and to ensure that examination results were appropriately evaluated and dispositioned. Qualification and certification records for examiners, equipment and consumables, and non-destructive examination procedures for the ISI activities were also reviewed.

Direct Observation:

Ultrasonic examinations (UT):

Visual Examination:

  • Steam Dryer Seismic Bracket @ 34°
  • Steam Dryer Seismic Bracket @ 146°
  • Steam Dryer Seismic Bracket @ 214°
  • Steam Dryer Seismic Bracket @ 326°
  • Steam Dryer Support Lugs to RPV @ 146°
  • Steam Dryer Support Lugs to RPV @ 214°
  • Steam Dryer Support Lugs to RPV @ 326° Automated UT:
  • Recirculation Pipe to Cross, Weld # 2B31-1IRCM-28BD-5 The inspectors also reviewed corrective action items such as Condition Reports (CRs)and Indication Notification Reports associated with the ISI program to determine if problems were being identified at appropriate thresholds and if adequate corrective actions were being taken. Documents reviewed are listed in the Attachment.

b. Findings

No findings of significance were identified.

1R11 Licensed Operator Requalification

a. Inspection Scope

Quarterly Resident Observation. The inspectors observed the performance of simulator scenario LT-SG-51071-00. The scenario involved activities including bypassing a failed local power range monitor, responding to a failed lube oil temperature controller on the 2A Reactor Feed Pump Turbine, Loss of Feedwater, Reactor Scram, RCIC turbine mechanical overspeed trip, failed governor on the HPCI turbine, and emergency depressurization. The inspectors reviewed procedures 10AC-MGR-019-0S, Procedure Use and Adherence, and DI-OPS-59-0896N, Operations Management Expectations, to assess operator performance for the following: formality of communication; procedure usage; alarm response; control board manipulations; group dynamics; and supervisory oversight. In addition, the inspectors reviewed the critique results from previous training sessions to assess performance improvement. The inspectors attended the licensees critique of operator performance to assess if the licensee-identified issues were comparable to issues identified by the inspectors. The inspectors compared their observations of licensee performance to the requirements in procedure DI-TRN-24-0885N, Simulator Documentation Requirements.

Requalification Examination Results Review. On October 15, 2004, the licensee completed the requalification annual operating tests required to be given to all licensed operators per 10 CFR 55.59(a)(2). The inspectors conducted an in-office review of the overall pass/fail results of the individual operating tests, and the crew simulator operating tests. These results were compared to the thresholds established in Manual Chapter 609 Appendix I, Operator Requalification Human Performance Significance Determination Process.

b. Findings

No findings of significance were identified.

1R12 Maintenance Effectiveness

a. Inspection Scope

The inspectors reviewed the following two maintenance activities associated with structures, systems, and components to assess the licensees implementation of the Maintenance Rule (10 CFR 50.65) with respect to the characterization of failures and the appropriateness of the associated (a)(1) or (a)(2) classification. The inspectors reviewed operator logs, associated CRs, Maintenance Work Orders, and the licensees procedures for implementing the Maintenance Rule. The review was to determine if equipment failures were being identified, properly assessed, and corrective actions established to return the equipment to a satisfactory condition. Documents reviewed are listed in the Attachment.

  • Failures in the CRD/Hydraulic Control Unit system

b. Findings

No findings of significance were identified.

1R13 Maintenance Risk Assessments and Emergent Work Evaluation

a. Inspection Scope

The inspectors reviewed the following five Plan of the Day (POD) documents listed below to verify that risk assessments were performed prior to components being removed from service. The inspectors reviewed risk assessment and risk management controls implemented for these activities to verify they were completed in accordance with procedure 90AC-OAM-002-0, Scheduling Maintenance, and 10 CFR 50.65 (a)(4).

For emergent work the inspectors assessed whether any increase in risk was promptly assessed and that appropriate risk management actions were implemented.

  • POD for the week of 1/22-28
  • POD for the week of 1/29-2/4
  • POD for the week of 2/5-11
  • POD for the week of 2/19-25
  • POD for the week of 3/12-18

b. Findings

No findings of significance were identified.

1R14 Personnel Performance During Non-Routine Plant Evolutions

a. Inspection Scope

For the three events described below, the inspectors observed operator actions and reviewed procedures, operator logs, computer data, and strip chart data recordings as applicable to verify plant responded as expected and that proper operator actions were taken. Documents reviewed are listed in the Attachment.

  • Unit 1 Shutdown for Drywell Cooler repairs on January 1
  • Unit 2 Shutdown for refueling on February 5
  • Notice of Unusual Event declared for a Freon leak in the Unit 2 Drywell Chiller system on February 7

b. Findings

No findings of significance were identified.

1R15 Operability Evaluations

a. Inspection Scope

The inspectors reviewed the following seven operability evaluations and compared the evaluations to the system requirements identified in the TS and the Final Safety Analysis Report (FSAR) to verify operability was adequately assessed and the system or component remained available to perform its intended function. Also, the inspectors assessed the adequacy of compensatory measures implemented as a result of the condition. Documents reviewed are listed in the Attachment.

  • CR 2005100077, Unit 1 Drywell Cooler Tube plugging
  • CR 2005101901, Control Room Door Latch Broken
  • CR 2005103163, Unit 2 Broken Torus T-quencher Bolts
  • CR 2005102226, 1C EDG Start Light Did Not Energize

b. Findings

No findings of significance were identified.

1R19 Post Maintenance Testing

a. Inspection Scope

The inspectors reviewed the procedures listed in the Attachment and observed personnel performance for the following seven maintenance and testing activities to verify procedural requirements were met. The inspectors also reviewed the activities to determine if the scope of testing demonstrated that the work performed was correctly completed and the affected equipment was functional and operable. Following the maintenance activities, the inspectors reviewed equipment status and alignment to verify the system or component was available to perform the required safety function.

  • 1C11C001A 1A CRD pump oil leaks
  • 2B21F037G Vacuum Breaker for SRV

b. Findings

No findings of significance were identified.

1R20 Refueling and Outage Activities

a. Inspection Scope

Unit 1 Drywell Cooler Repair. The inspectors reviewed the outage schedule to verify the licensees use of risk management techniques, incorporation of operating experience, and past lessons learned for the scheduled maintenance outage conducted January 3 -

January 10. Additionally, the inspectors routinely reviewed procedure DI-OPS-57-0393, Outage Safety Assessment, to verify the licensee was correctly maintaining required equipment. During the planned outage, the inspectors monitored licensee control over the following outage activities listed below.

  • Walkdown of the drywell to verify material conditions supported plant operations
  • Plant startup, heatup, and power ascension
  • Licensee identification and resolution of problems related to the planned outage activities Unit 2 Refueling Outage. The inspectors reviewed procedure DI-OPS-57-0393 and the outage schedule to verify the licensees use of risk management techniques, incorporation of operating experience, and past lessons learned for the refueling outage conducted February 4 - March 14. Additionally, the inspectors reviewed the outage safety assessment to verify the licensee had contingency plans and these plans included sufficient equipment to maintain a defense-in-depth approach to safety. The inspectors routinely reviewed procedure DI-OPS-57-0393 to verify the licensee was correctly maintaining required equipment in service in accordance with the overall outage safety assessment. During the refueling outage, the inspectors monitored licensee control over the outage activities listed below. Documents reviewed are listed in the Attachment.
  • Four clearances to verify implementation of the clearance process and the associated equipment was properly configured to support the function of the clearance
  • Calibration of reactor instrumentation used to monitor reactor water level within surveillance requirements
  • Observe fuel movement during initial and final fuel shuffles
  • TS and procedures to verify mode change requirements were met
  • Walkdown of the drywell and torus proper and other areas to verify material conditions supported plant operations
  • Plant startup, heatup, and power ascension
  • Licensee identification and resolution of problems related to outage activities

b. Findings

No findings of significance were identified.

1R22 Surveillance Testing

a. Inspection Scope

The inspectors reviewed surveillance test procedures and either witnessed the test or reviewed test records for the following six surveillances to verify the test adequately demonstrated that the affected equipment was operable. The inspectors reviewed the activities to assess for preconditioning of equipment, procedure adherence, and equipment alignment following completion of the surveillance. The inspectors reviewed procedure AG-MGR-21-0386N, Evolution and Pre-and Post-Job Brief Guidance, and attended selected briefings to verify procedural requirements were met.

Surveillance Tests

Runback Test In-Service Tests

b. Findings

No findings of significance were identified.

1R23 Temporary Plant Modifications

a. Inspection Scope

The inspectors reviewed the following two temporary modifications (TMM) and assessed each evaluation using criteria defined in procedure 40AC-ENG-018-0S, Temporary Modification Control. In addition, the 10 CFR 50.59 evaluations were assessed using the design basis information provided in the FSAR to verify the modifications did not affect the safety functions of these systems. The inspectors checked the modifications to verify they were installed in accordance with the TMM requirements. Documents reviewed are listed in the Attachment.

  • TMM 1-05-001, Installation of a PSW Blank on Cooler 1T47-B009B
  • TMM 2-05-001, Temporary Power to Unit 2 Drywell Chiller System Freon Monitor

b. Findings

No findings of significance were identified.

RADIATION SAFETY

Cornerstones: Occupational Radiation Safety and Public Radiation Safety 2OS1 Access Controls To Radiologically Significant Areas Access Controls. The inspectors reviewed and evaluated licensee guidance and its implementation for controlling and monitoring worker access to radiologically significant areas and tasks associated with the refueling outage. The inspectors evaluated changes to, and adequacy of procedural guidance; directly observed implementation of established administrative and physical radiological controls; appraised radiation worker and technician knowledge of, and proficiency in implementing radiation protection activities; and assessed radiation worker (radworker) exposures to radiation and radioactive material.

The inspectors directly observed controls established for workers and Health Physics Technician (HPT) staff in potential airborne radioactivity area, radiation area, high radiation area (HRA), locked-high radiation area (LHRA), and very high radiation area (VHRA) locations. Controls and their implementation for LHRA keys and for storage of irradiated material within the spent fuel pool locations were reviewed and discussed in detail. Established radiological controls were evaluated for selected outage tasks including insulation removal/replacement; ISI; fuel movement; local power range monitor (LPRM) replacement; CRDM removal, replacement, and shipping; and radioactive waste (radwaste) processing, storage, and shipping. In addition, licensee controls for areas where dose rates could change significantly as a result of plant shutdown and refueling operations were reviewed and discussed.

For selected tasks, the inspectors attended pre-job briefings and reviewed radiation work permit (RWP) details to assess communication of radiological control requirements to workers. Occupational workers adherence to selected RWPs and HPT proficiency in providing job coverage were evaluated through direct observations and interviews with licensee staff. Electronic dosimeter (ED) alarm setpoints and worker stay times were evaluated against area radiation survey results for ISI activities, LPRM replacement, and CRDM removal and replacement. Worker exposure as measured by ED and by licensee evaluations of skin doses resulting from discrete radioactive particle or dispersed skin contamination events during current outage activities were reviewed and assessed independently. For HRA tasks involving significant dose gradients, e.g., ISI, CRDM removal/replacement, and LPRM replacement, the inspectors evaluated the use and placement of whole body and extremity dosimetry to monitor worker exposure.

Postings and physical controls established within the radiologically controlled area for access to the Unit 2 drywell and torus; Unit 1 and Unit 2 reactor and turbine building locations; Unit 1 and Unit 2 radioactive waste processing, storage, shipping equipment, and locations; and the low level waste sorting facilities were evaluated during facility tours. The inspectors independently measured radiation dose rates or directly observed conduct of licensee radiation surveys and results for ISI tasks, CRDM removal/replacement, and radioactive material/waste shipping tasks. Results were compared to current licensee surveys and assessed against established postings and radiation controls. Licensee access controls were observed for selected Unit 1 and Unit 2 LHRA and VHRA locations.

The inspectors evaluated implementation and effectiveness of licensee controls for both airborne and external radiation exposure. The inspectors reviewed and discussed selected whole-body count analyses conducted between January 1, 2004, and March 3, 2005. This was performed to evaluate the implementation and effectiveness of personnel monitoring and administrative and physical controls including air sampling, barrier integrity, engineering controls, and postings for tasks having the potential for individual worker internal exposures to exceed 30 millirem Committed Effective Dose Equivalent. Effectiveness of external radiation exposure controls were evaluated through review and discussions of individual worker dose as measured by ED for selected tasks.

Radiation protection activities were evaluated against the FSAR, the TS, and 10 CFR Parts 19 and 20 requirements. Specific assessment criteria included FSAR Section 11, Radioactive Waste Management, and Section 12, Radiation Protection; 10 CFR 19.12; 10 CFR 20, Subpart B, Subpart C, Subpart F, Subpart G, Subpart H, and Subpart J; TS Sections 5.4.1(a), Procedures, and 5.7, High Radiation Area; and approved procedures.

Documents and records reviewed are listed in the Attachment.

Problem Identification and Resolution. Licensee Corrective Action Program (CAP)documents associated with access control to radiologically significant areas were reviewed and assessed. The inspectors evaluated the licensees ability to identify, characterize, prioritize, and resolve the identified issues in accordance with procedure NMP-GM-002-GL02, Corrective Action Program Details and Expectations Guideline.

Documents reviewed are listed in the Attachment.

b. Findings

No findings of significance were identified.

2OS2 As Low As Reasonably Achievable (ALARA) Planning and Controls

a. Inspection Scope

ALARA. The inspectors reviewed ALARA program guidance and its implementation for ongoing outage job tasks. The inspectors evaluated the accuracy of ALARA work planning and dose budgeting, observed implementation of ALARA initiatives and radiation controls for selected jobs in-progress, assessed the effectiveness of source-term reduction efforts, and reviewed historical dose information.

ALARA planning documents and procedural guidance were reviewed and projected dose estimates were compared to actual dose expenditures for the following high dose jobs: ISI work, CRDM removal, and LPRM replacement activities. Differences between budgeted dose and actual exposure received were discussed with cognizant ALARA staff. Changes to dose budgets relative to changes in radiation source term and/or job scope were also discussed. The inspectors attended pre-job briefings and evaluated the communication of ALARA goals, RWP requirements, and industry lessons-learned to job crew personnel. The inspectors also attended three Plant ALARA Review Committee meetings and observed the interface between plant management and ALARA planning staff.

The inspectors made direct field or closed-circuit-video observations of outage job tasks involving ISI activities and CRDM removal/replacement. For the selected tasks, the inspectors evaluated radworker and HPT job performance; individual and collective dose expenditure versus percentage of job completion; surveys of the work areas, appropriateness of RWP requirements; and adequacy of implemented engineering controls. For ISI work, the inspectors interviewed radworkers and job sponsors regarding understanding of dose reduction initiatives and their current and expected accumulated doses at completion of the job tasks.

Implementation and effectiveness of selected program initiatives with respect to source-term reduction were evaluated. Chemistry program ALARA initiatives and their effect on Unit 2 drywell dose rate trends were reviewed. The effectiveness of temporary shielding installed for the current outage was assessed through review of shielding request packages and pre-shielding versus post-shielding dose rate data. The inspectors also reviewed the results of pipe flushing to reduce Unit 2 reactor nozzle dose rates.

Plant exposure history for 2001 through 2003 and data reported to the NRC pursuant to 10 CFR 20.2206 were reviewed as were established goals for reducing collective exposure during the refueling outage. The inspectors reviewed procedural guidance for dosimetry issuance and exposure tracking. The inspectors also examined dose records of declared pregnant workers to evaluate assignment of gestation dose. In addition, selected individual access records were reviewed for dose received during work in areas with high dose rate gradients.

ALARA program activities and their implementation were reviewed against 10 CFR Part 20, and approved procedures. In addition, licensee performance was evaluated against guidance contained in Regulatory Guide (RG) 8.8, Information Relevant to Ensuring that Occupational Radiation Exposures at Nuclear Power Stations will be As Low As Reasonably Achievable, and RG 8.13, Instruction Concerning Prenatal Radiation Exposure. Documents reviewed are listed in the Attachment.

Problem Identification and Resolution. The inspectors reviewed five CRs, two self-assessments, and one corporate audit in the area of exposure control. The inspectors evaluated the licensees ability to identify, characterize, prioritize, and resolve the identified issues in accordance with procedure NMP-GM-002-GL02. Documents reviewed are listed in the Attachment.

b. Findings

No findings of significance were identified.

2PS2 Radioactive Material Processing and Transportation

a. Inspection Scope

Waste Processing and Characterization. The inspectors reviewed and discussed the currently installed radwaste processing systems as described in the FSAR, Section 11.

In addition, stored and disposed radwaste types and quantities as documented in Effluent Release Report for 2004 were discussed with responsible licensee representatives.

The operability and configuration of selected liquid and solid radioactive radwaste processing systems and equipment were evaluated. Inspection activities included document review, interviews with plant personnel, and direct inspection of processing equipment and piping. The inspectors directly observed equipment material condition and configuration for liquid and solid radwaste processing systems and licensee staff were interviewed regarding equipment function and operability. The licensees policy regarding abandoned radwaste equipment was reviewed and discussed with cognizant licensee representatives. Operations staff were interviewed to assess knowledge of radwaste system processing operations. Procedural guidance involving resin dewatering activities and filling of waste packages was reviewed for consistency with the licensees Process Control Program and FSAR details.

Licensee radionuclide characterizations of each major waste stream were evaluated.

For dry active waste (DAW), primary resin, secondary resin, and filters, the inspectors evaluated PCP and licensee procedural guidance against 10 CFR 61.55 and the Branch Technical Position on Radioactive Waste Classification details. Part 61 data and scaling factors were reviewed and discussed with licensee representatives for radwaste processed or transferred to licensed burial facilities for the period January 1, 2003, through January 31, 2005. The licensees analyses and current scaling factors for quantifying hard-to-detect nuclides were assessed. The inspectors discussed potential for changes plant operating conditions and reviewed selected DAW waste stream radionuclide data to determine if known plant changes were assessed and radionuclide composition remained consistent for the period reviewed. Documents reviewed are listed in the Attachment.

Transportation. The inspectors evaluated licensee activities related to transportation of radioactive material. The evaluation included review of shipping records and procedures, assessment of worker training and proficiency, and direct observation of shipping activities.

The inspectors assessed shipping-related procedures for compliance to applicable regulatory requirements. Selected shipping records were reviewed for completeness and accuracy, and for consistency with procedures. Training records for individuals qualified to ship radioactive material were checked for completeness. In addition, specific training curricula provided to maintenance workers were assessed. For a shipment of radioactive waste resin to a processing vendor and a shipment of CRDM equipment to a refurbishment facility, the inspectors directly observed initial package preparation; independently verified results of contamination and direct radiation surveys; evaluated shipping paperwork for completeness; and assessed initial loading, bracing, and placarding of the transport vehicles. Responsible staff were interviewed to assess their knowledge of package preparation specifications, and applicable radiation and contamination control limits.

Transportation program guidance and implementation were reviewed against regulations detailed in 10 CFR 71, and 49 CFR 170-189 and applicable procedures. In addition, training activities were assessed against 49 CFR 172 Subpart H, and the guidance documented in NRC Bulletin 79-19. Documents reviewed are listed in the

.

Problem Identification and Resolution. Licensee CAP documents associated with radwaste processing and transportation activities were reviewed and assessed. The inspectors evaluated the licensees ability to identify, characterize, prioritize, and resolve the identified issues in accordance with procedure NMP-GM-002-GL02. Documents reviewed are listed in the Attachment.

b. Findings

Introduction.

A Green NRC-identified non-cited violation (NCV) of 10 CFR 71.5 was identified for failure to implement current package design specifications for proper closing of Type A shipping packages (CRDM shipment boxes) as required by DOT regulations. Specifically, for Type A packages containing CRDM equipment shipped between January 2003 and February 2005, the licensee failed to prepare the package closures in accordance with vendor package specifications as required by 49 CFR 173.475(e).

Description.

The inspectors identified significant differences in selected package certification details, design/testing specifications, and associated drawings referenced within container certification and engineering evaluations maintained by the licensee for use in preparation of CRDM equipment DOT Type A packages (Model IAEA-102.2-3.5-7A-TRF). These Type A packages were being used for shipment of CRDM equipment.

Subsequent calls to the vendor supplying the Type A packages resulted in receipt of current and accurate package certification and engineering evaluation documentation.

The inspectors noted that the documents provided required package preparation specifications including package maximum weight and lid closure device (T-bolt) torque values and configuration requirements, i.e., required types and numbers, and conditions of the washers used for final lid closure assembly. From review and discussion of quality control documents associated with previous Type A package CRDM equipment shipments made between January 2003 and February 2005, the inspectors noted that required lid closure device torque values and assembly configurations were not specified. Licensee representatives stated that previous package preparation guidance did not specify a required closure torque value nor a closure device configuration, but only required the verification that the T-bolts and lid were secured tightly. After this issue was identified, the inspectors observed subsequent shipment packaging for CRDM equipment and verified that the licensees preparation met the current and approved engineering document specifications.

Analysis.

This finding is more than minor because it is associated with the public radiation cornerstone program and process attribute and it affected the cornerstone objective to ensure adequate protection of public health and safety from exposure to radioactive material released into the public domain. The issue was reviewed using the Public Radiation Safety Significance Determination Process and was determined to be of very low safety significance (Green) because a radiation limit was not exceeded nor was the packaging breached. In addition, previous shipments made by the licensee had arrived at their destination with no identified degradation of the subject packaging and immediate corrective actions assured that on-going CRDM equipment packages were prepared properly prior to shipment.

Enforcement.

10 CFR 71.5 requires licensees to conform with the regulations in DOT 49 CFR Parts 170 through 189. For Type A package shipments, 49 CFR 173.415(a)requires each offeror of a Specification 7A Type A package to maintain complete documentation of tests and engineering evaluation or comparative data showing the construction methods, packaging design and materials of construction to comply with that specification for at least one year after the latest shipment. Further, 49 CFR 173.475(e) requires that each special instruction for closing and preparation of a package be followed. For shipments of Type A CRDM packages made between January 2003 and February 2005, the licensee failed to implement current design document specifications for closure of the DOT Type 7A packages, in that, T-bolt torque values and closure assembly specifications for package closure were not met. The licensee documented this issue in its CAP as CR No. 2005101950. Since this violation is of very low safety significance and the licensee entered the finding into its CAP, this violation is being treated as a NCV consistent with Section VI.A of the NRC Enforcement Policy: NCV 05000321, 366/2005002-01, Failure to Implement Appropriate DOT Type A Package Closure Requirements.

OTHER ACTIVITIES

4OA1 Performance Indicator (PI) Verification

a. Inspection Scope

The inspectors sampled licensee data for the PIs listed below. To verify the accuracy of the PI data reported, PI definitions and guidance contained in NEI 99-02, "Regulatory Assessment Indicator Guideline," Rev. 2, were used to screen each data element.

Public Radiation Safety Cornerstone

  • RETS/ODCM Effluent Occurrence The inspectors reviewed the PI results for the period April 2004 through December 2004. For the assessment period, the inspectors reviewed cumulative and projected doses to the public, out-of-service effluent radiation monitors and selected compensatory sampling data, and two CRs related to RETS/ODCM issues. The inspectors also reviewed licensee procedural guidance for collecting and documenting PI data. Documents reviewed are listed in the Attachment.

Occupational Radiation Safety Cornerstone

  • Occupational Exposure Control Effectiveness The inspectors reviewed the PI results for the period from April 2004 through December, 2004. For the assessment period, the inspectors reviewed electronic dosimeter alarm logs and two CRs related to controls for exposure significant areas. The inspectors also reviewed licensee procedural guidance for collecting and documenting PI data.

Documents reviewed are listed in the Attachment.

b. Findings

No findings of significance were identified.

4OA2 Identification and Resolution of Problems

1. Daily Condition Report Review

a. Inspection Scope

As required by Inspection Procedure 71152, Identification and Resolution of Problems, and in order to help identify repetitive equipment failures or specific human performance issues for follow-up, the inspectors performed a daily screening of items entered into the licensees corrective action program. This review was accomplished by reviewing hard copies of each condition report and accessing the licensees computerized database.

2. Annual Sample Review

a. Inspection Scope

The inspectors performed detailed reviews of the following two CRs to verify the full extent of the issue was identified, an appropriate evaluation was performed, and appropriate corrective actions were specified and prioritized. The inspectors evaluated the CR against the licensees corrective action program as delineated in procedure NMP-GM-002, Corrective Action Program, and 10 CFR 50, Appendix B.

  • CR 2005100077, Increased Unit 1 Drywell Floor Drain Leakage

b. Findings and Observations

No findings of significance were identified. The inspectors noted that the CRs were timely and accurately described the condition, the cause evaluations were thorough, and the corrective actions appeared to address the identified causes and were appropriately scheduled.

4OA5 Other Activities

1. Independent Spent Fuel Storage Installation (ISFSI) Radiological Controls

a. Inspection Scope

The inspectors observed gamma-ray, neutron, and contamination surveys of the ISFSI facility and compared the results to previous monthly surveys and TS limits. The inspectors also observed and evaluated implementation of radiological controls, including RWPs and postings, and discussed the controls with an HPT and health physics supervisory staff. One CR and one procedure for ISFSI radiological controls were also reviewed and discussed. Radiological control activities for ISFSI areas were evaluated against 10 CFR Part 20, 10 CFR Part 72, ISFSI TS, and ISFSI Certificate of Compliance. Documents reviewed are listed in the Attachment.

b. Findings

No findings of significance were identified.

2. Temporary Instruction (TI) Status

The following TIs were previously documented in the associated integrated inspection report (IIR). This section is provided to administratively update the TI status stated in the IIR with the current status. This completes the regional inspection requirements for these TIs.

TI Number IIR Number IIR Status Current Status 2515/154 05000321, 05000366/2004002 Discussed Closed 2515/156 05000321, 05000366/2004003 Discussed Closed

3. Task Interface Agreement (TIA) 2005-003 Related to the Operability and Validity of the

Local Leak Rate Testing of the Main Steam Isolation Valves - Plant Edwin I. Hatch During local leakrate testing of main steam isolation vavles, the residents determined there were additional questions concerning the local leak rate testing performed. NRC Region II has requested technical assistance from the office of Nuclear Reactor Regulation via TIA 2005-003. This TIA is available from the NRCs document system (ADAMS) as accession number ML051160420. ADAMS is accessible from the NRC Web site at http://www.nrc.gov/reading-rm/adams.html.

4OA6 Meetings, Including Exit

1.

Exit Meeting Summary

On April 7, the inspectors presented the inspection results to Mr. Dennis Madison and the other members of his staff who acknowledged the findings. The inspectors confirmed that appropriate controls were implemented for proprietary information provided to the resident office during the inspection.

2. Annual Assessment Meeting Summary

On April 13, the NRCs Chief of Reactor Projects Branch 2 and Resident Inspector assigned to the Edwin I. Hatch Nuclear Plant (HNP) met with Southern Nuclear Operating Company to discuss the NRCs Reactor Oversight Process (ROP) and the NRCs annual assessment of HNP safety performance for the period of January 1, 2004 -

December 31, 2004. The major topics addressed were: the NRCs assessment program and the results of the HNP assessment. This meeting was open to the public. A listing of meeting attendees and information presented during the meeting are available from the NRCs document system (ADAMS) as accession number ML051150126. ADAMS is accessible from the NRC Web site at http://www.nrc.gov/reading-rm/adams.html.

ATTACHMENT:

SUPPLEMENTAL INFORMATION

KEY POINTS OF CONTACT

Licensee personnel

R. Dedrickson, Assistant General Manager - Plant Support
J. Dixon, Health Physics Manager
G. Frederick, General Manager - Nuclear Plant
M. Googe, Maintenance Manager
J. Hammonds, Operations Manager
J. Lewis, Training and Emergency Preparedness Manager
D. Madison, Assistant General Manager - Plant Operations
J. Thompson, Nuclear Security Manager
R. Varnadore, Engineering Manager

LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED

Closed

2515/154 (Units 1 and 2) TI Spent Fuel Material Control and Accounting at Nuclear Power Plants (Section 4OA5.2)

2515/156 (Units 1 and 2) TI Offsite Power System Operational Readiness (Section 4OA5.2)

Opened and Closed

05000321, 366/2005002-01 NCV Failure to Implement Appropriate DOT Type A Package Closure Requirements (Section 2PS2)

LIST OF DOCUMENTS REVIEWED

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