IR 05000302/2007005

IR 05000302-07-005; on 10/01- 12/31/2007; Crystal River, Unit 3; NRC Integrated Inspection Report

ML080350436
Person / Time
Site:	Crystal River
Issue date:	01/29/2008
From:	Vias S NRC/RGN-II/DRP/RPB3
To:	Young D Florida Power & Light Co
References
FOIA/PA-2010-0209 IR-07-005
Download: ML080350436 (43)

Text

ary 29, 2008

SUBJECT:

CRYSTAL RIVER UNIT 3 - NRC INTEGRATED INSPECTION REPORT 05000302/2007005

Dear Mr. Young:

On December 31, 2007, the US Nuclear Regulatory Commission (NRC) completed an inspection at your Crystal River Unit 3. The enclosed integrated inspection report documents the inspection findings, which were discussed on January 7, 2008, with you and members of your staff.

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

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

Based on the results of this inspection, the inspectors identified one finding of very low safety significance (Green). The finding was determined to involve a violation of NRC requirements.

However, because of the very low safety significance of the issue, and because it was entered into your corrective action program, the NRC is treating the issue as a non-cited violation (NCV)

consistent with Section VI.A of the NRC Enforcement Policy. If you contest this non-cited violation, you should provide a response within 30 days of the date of this inspection report, with the basis for your denial to the Nuclear Regulatory Commission, ATTN.: Document Control Desk, Washington DC 20555-0001; with copies to the Regional Administrator, NRC Region II; The Director, Office of Enforcement, United States Nuclear Regulatory Commission, Washington DC 20555-0001; and the NRC Resident Inspector at the Crystal River Unit 3 site.

FPC 2 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 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 By M. Sykes For/

Steven J. Vias, Chief Reactor Projects Branch 3 Division of Reactor Projects Docket No.: 50-302 License No.: DPR-72

Enclosure:

Inspection Report 05000302/2007005 w/Attachment: Supplemental Information

REGION II==

Docket No.: 50-302 License No.: DPR-72 Report No: 05000302/2007005 Licensee: Progress Energy Florida (Florida Power Corporation)

Facility Crystal River Unit 3 Location: 15760 West Power Line Street Crystal River, FL 34428-6708 Dates: October 1, 2007 - December 31, 2007 Inspectors: T. Morrissey, Senior Resident Inspector R. Reyes, Resident Inspector R. Aiello, Senior Operations Engineer (Section 1R11)

R. Carrion, Senior Reactor Inspector (Section 1R08)

J. Fuller, Reactor Inspector (Section 1R08)

C. Peabody, Reactor Inspector (Section 4OA5)

A. Nielsen, Health Physicist (Sections 2OS2 and 4OA1)

G. Kuzo, Senior Health Physicist (Sections 2PS1 and 2PS2)

R. Hamilton, Senior Health Physicist (Sections 2OS1 and 4OA1)

Approved by: Steven J. Vias, Chief Reactor Projects Branch 3 Division of Reactor Projects Enclosure

SUMMARY OF FINDINGS

IR 05000302/2007-005; 10/01/2007 - 12/31/2007; Crystal River Unit 3; Refueling and Other

Outage Activities.

The report covered a three-month period of inspection by the resident inspectors. One Green non-cited violation was identified during this inspection. The significance of most findings is indicated by their color (Green, White, Yellow, 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 NRC 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 4, dated December 2006.

Cornerstone: Mitigating Systems

Green.

The inspectors identified a non-cited violation (NCV) of Improved Technical Specification 5.6.1.1.a, for failure to adequately implement procedures required by Regulatory Guide 1.33, Appendix A, Section 3, Procedures for Startup, Operation, and Shutdown of Safety-Related PWR Systems. Specifically, the licensee failed to verify no latent debris was present in containment that could impact the emergency core cooling system (ECCS) sump. Corrective actions completed include: removal of the debris identified by the inspectors and performing additional inspection and cleaning of containment.

The finding is more than minor because it could be reasonably viewed as a precursor to a significant event involving debris accumulation on the containment sump screens which could cause impairment to ECCS recirculation flow during a design basis loss of coolant accident. The inspectors referenced Inspection Manual Chapter 0609,

Significance Determination Process (SDP), Phase 1 screening and determined the finding to be of very low safety significance. Although the debris impacted the mitigating system cornerstone, it was unlikely to have resulted in an actual loss of safety function and was not potentially risk significant due to possible external events. A contributing cause of this finding is related to the crosscutting area of Human Performance, specifically Work Practices in that the licensee did not adequately comply with a containment inspection procedure. (IMC 305, H.4(b))

B. Licensee-identified Violations None

REPORT DETAILS

Summary of Plant Status:

The unit operated at essentially 100 percent rated thermal power (RTP) until October 29 when power was reduced to approximately 61 percent RTP after condensate pump CDP-1B was secured due to a problem with its control circuit. The unit was shut down for a planned refueling outage on November 3rd. The unit was restarted on December 6th and resumed 100 percent RTP on December 9th. The unit operated at essentially 100 percent RTP for the remainder of the inspection period.

REACTOR SAFETY

Cornerstones: Initiating Events, Mitigating Systems, Barrier Integrity

1R01 Adverse Weather Protection

Seasonal Susceptibility: Cold Weather Preparation

a. Inspection Scope

The inspectors evaluated the licensees readiness for mitigating cold weather to assure that vital systems and components were protected from freezing in accordance with the licensees Administrative Instruction AI-513, Seasonal Weather Preparations, Section 4.1, Cold Weather Preparations. The inspectors walked down portions of the systems/areas listed below to check for any unidentified susceptibilities. Operability of heat trace circuits was verified. Nuclear condition reports (NCRs) were reviewed to check that the licensee was identifying and correcting cold weather protection issues.

• Emergency feedwater pump (EFP)-3 building

• Emergency diesel generator (EGDG) rooms

• Raw water and service water systems room There were no sustained periods of freezing weather during the inspection period.

b. Findings

No findings of significance were identified.

1R04 Equipment Alignment

Partial System Walkdowns

a. Inspection Scope

The inspectors performed walkdowns of the critical portions of the selected trains to verify correct system alignment. The inspectors reviewed plant documents to determine the correct system and power alignments, and the required positions of select valves and breakers. The inspectors verified that the licensee had properly identified and resolved equipment alignment problems that could cause initiating events or impact mitigating system availability. The inspectors verified the following three partial system alignments in system walkdowns using the listed documents:

C Raw water pump (RWP)-3A, using operating procedure (OP)-408, Nuclear Services Cooling System; and A train decay heat removal (DHR) systems, using OP-404, Decay Heat Removal System, while RWP-3B was out of service for testing, C EGDG-1A and its associated 4160V engineered safeguards (ES) bus, using OP-707, Operation of the Engineered Safeguards Diesel Generators, and surveillance procedure (SP)-321, Power Distribution Breaker Alignment and Power Availability Verification, while the EGDG-1B system was out of service for maintenance, and C A and B trains spent fuel pool cooling, service water pump (SWP)-1B and RWP-2B systems utilizing OP-406, Spent Fuel Cooling System and OP-408, Nuclear Services Cooling System, with a full core off load in the spent fuel pool.

b. Findings

No findings of significance were identified.

1R05 Fire Protection

Fire Protection Walkdowns

a. Inspection Scope

The inspectors walked down accessible portions of the plant to assess the licensees implementation of the fire protection program. The inspectors checked that the areas were free of transient combustible material and other ignition sources. Also, fire detection and suppression capabilities, fire barriers, and compensatory measures for fire protection problems were verified. The inspectors checked fire suppression and detection equipment to determine whether conditions or deficiencies existed which could impair the function of the equipment. The inspectors selected the areas based on a review of the licensees probabilistic risk assessment and ongoing work activities. The inspectors also reviewed the licensees fire protection program to verify the requirements of Final Safety Analysis Report (FSAR) Section 9.8, Plant Fire Protection Program, were met. Documents reviewed are listed in the attachment. The inspectors walked down the following nine areas important to reactor safety:

• Control complex chiller and ventilation room

• Emergency feedwater initiation and control (EFIC) rooms

• EFP-3 building

• Intermediate building, main steam isolation and atmosphere dump valve areas

• B train decay heat removal (DHR) and building spray (BS) vault

• Emergency diesel generator EGDG-1C building

• Auxiliary building fuel handling floor

• Reactor building - top of pressurizer area

• Main control room

b. Findings

No findings of significance were identified.

1R06 Flood Protection Measures

.1 Internal Flood Protection

a. Inspection Scope

The inspectors reviewed the Crystal River Unit 3, Final Safety Analysis Report (FSAR), Chapter 2.4.2.4, Facilities Required for Flood Protection, and the Crystal River Unit 3 Design Basis Documents that depicted protection for areas containing safety-related equipment to identify areas that may be affected by internal flooding. A walkdown of emergency feed pump EFP-1 and EFP-2 area was conducted to ensure that flood protection measures were in accordance with design specifications. Specific plant attributes that were checked included structural integrity, sealing of penetrations, and operability of sump systems.

b. Findings

No findings of significance were identified.

1R08 Inservice Inspection (ISI) Activities

.1 Inservice Inspection Activities Other than Steam Generator Tube Inspections, PWR

Vessel Upper Head Penetration Inspections, and Boric Acid Corrosion Control

a. Inspection Scope

From November 13 - November 20, 2007, the inspectors reviewed the implementation of the licensees ISI program for monitoring degradation of the reactor coolant system (RCS) boundary and other risk significant piping system boundaries for Crystal River Unit 3. The inspectors selected a sample of American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code,Section XI required examinations for review.

The inspectors conducted an on-site review of nondestructive examination (NDE)activities to evaluate compliance with Technical Specifications and the applicable editions of ASME Section V and XI (1989 Edition/No Addenda) and to verify that indications and defects (if present) were appropriately evaluated and dispositioned in accordance with the requirements of ASME Section XI acceptance standards.

Specifically, the inspectors directly observed the NDE activities described below and reviewed the corresponding NDE procedures, NDE reports, equipment and consumables certification records, and personnel qualification records.

• Computer Radiographic Examination (RT) of weld FW-00-029 R1 (Repair 1), 18 Main Feed Water, ASME Class 2

• Computer RT of weld FW-00-028 R1 (Repair 1), 18 Main Feed Water, ASME Class 2

• Ultrasonic Examination (UT) of weld B1.1.1, Nozzle Belt to Upper Shell Weld, Reactor Pressure Vessel Internal Weld, ASME Class 1

• Magnetic Particle Examination (MT) of weld FW-00-028 R1 (Repair 1), 18 Main Feed Water, ASME Class 2

• Eddy Current Testing (ECT) of A and B Steam Generator tubes, ASME Class 1 (sample of data acquisition and resolution analysis)

Specifically, the inspectors reviewed the following examination records and the associated NDE procedures, equipment and consumables certification records, and personnel qualification records.

• Computer RT of welds FW-00-028, FW-00-028 R2 (Repair 2), FW-00-029, and FW-00-029 R2 (repair 2), 18 Main Feed Water, ASME Class 2

• Visual Examination Leak Test, VT-07-164, RCT-1 Vessel Manway Flange, ASME Class 1 Bolted Connection

• Visual Examination Leak Test, VT-07-178, RCRE-1 Vessel Head Flange, ASME Class 1 Bolted Connection

• Visual Examination Leak Test, VT-07-161, RCV-10 Valve Flange Connection Bolting, ASME Class 1 Bolted Connection

• Visual Examination Leak Test, VT-04-014, EFP-3 Buried Piping Pressure Test, ASME Class 3

• Visual Examination Leak Test, VT-07-026, EFP-3 Buried Piping Pressure Test, ASME Class 3

• Visual Examination for Boric Acid Detection, VT-07-162, Reactor Pressure Vessel Head and 69 Control Rod Drive Mechanisms

• UT-07-040, Makeup and Purification System Pipe to Elbow Weld, ASME Class 2 The inspectors reviewed the following NDE reports with recordable indications to ensure that they were properly dispositioned in accordance with the applicable ASME Section XI acceptance criteria.

• Report Number VT-05-170, VT-3 Examination of Liner Plate Penetration 426

• Report Number VT-05-209, VT-3 Examination of Liner Plate Penetration 426

• Report Number VT-07-002, VT-3 Examination of Support DHH-558

• Report Number VT-07-013, VT-3 Examination of Support DHR-38 The inspectors observed and reviewed in-process welding activities performed during this outage. Specifically, the inspectors observed welding on Alloy 600 pressurizer weld overlays and Alloy 600 pressurizer nozzle repairs. The inspectors reviewed welding records and observed in-process post-weld heat treatment for FW-00-028 and FW-00-029, main feedwater piping welds. The inspectors reviewed welding procedures, procedure qualification records, welder qualification records, and NDE reports for the welds.

b. Findings

No findings of significance were identified.

.2 Boric Acid Corrosion Control Inspection Activities

a. Inspection Scope

The inspectors reviewed the licensees Boric Acid Corrosion Control Program (BACCP)to ensure compliance with commitments made in response to NRC Generic Letter 88-05, Boric Acid Corrosion of Carbon Steel Reactor Pressure Boundary, and Bulletin 2002-01, Reactor Pressure Vessel Head Degradation and Reactor Coolant Pressure Boundary Integrity.

The inspectors conducted an on-site record review, and an independent walk-down of the auxiliary building and the containment building, which is not normally accessible during at-power operations, to evaluate licensee compliance with their program procedures and applicable industry guidance. In particular, the inspectors verified that the licensees visual examinations focused on locations where boric acid leaks could cause degradation of safety-significant components and that degraded or non-conforming conditions were properly identified in the licensees corrective action program.

The inspectors reviewed the results of the Refueling Outage 15 ASME Class 1 Bolted Connection Inspection as required by IWA-5242 of ASME Section XI, 1989 Edition with no Addenda, and the licensees boric acid corrosion control program. The inspectors also reviewed the results of the last ASME Section XI, Class 1 System Leakage Test.

The inspectors reviewed the results of the licensees Mode 3 walkdowns completed March 18, 2006, August 19, 2006, and February 22, 2007. These walkdowns were prescribed by the BACCP, but were not required by ASME Section XI.

The inspectors reviewed a sample of engineering evaluations completed for boric acid found on reactor coolant system piping and/or other ASME code class components to verify that the minimum design code required section thickness had been maintained for affected components. The inspectors also reviewed licensee corrective action documents initiated for evidence of boric acid leakage to confirm that those documents were consistent with the requirements of Section XI of the ASME Code, 10 CFR 50 Appendix B Criterion XVI, and licensee BACCP procedures.

b. Findings

No findings of significance were identified.

.3 Steam Generator (SG) Tube Inspection Activities

a. Inspection Scope

From November 13 - 20, 2007, the inspectors reviewed the Unit 3 SG tube ECT examination activities to ensure compliance with Technical Specifications (TS),applicable industry operating experience and technical guidance documents, and ASME Code Section XI requirements.

The inspectors reviewed licensee SG inspection activities to ensure that ECT inspections were conducted in accordance with the licensees SG Program and applicable industry standards. The inspectors reviewed the SG examination scope, ECT acquisition procedures, site-specific Examination Technique Specification Sheets, the most recent SG degradation assessment, and the last condition monitoring and operational assessment. The inspectors reviewed documentation to ensure that the ECT probes and equipment configurations used were qualified to detect the expected types of SG tube degradation, and a sampling of tube data was reviewed with the independent qualified data analyst. The inspectors also verified that the appropriate inspection scope expansion criteria was applied based on inspection results. The licensee identified a number of tubes that were required to be removed from service, and the inspectors reviewed the licensees tube plugging criteria to verify that it had been appropriately applied. The inspectors ensured that all tubes with relevant indications were appropriately screened for in-situ pressure testing. The licensee identified that three tubes meet the screening criteria for in-situ pressure testing. The NRC inspectors reviewed the in-situ pressure testing plans and the resulting pressure versus time traces. Additionally, the inspectors observed in-situ pressure testing activities for one tube.

b. Findings

No findings of significance were identified.

.4 Identification and Resolution of Problems

The inspectors performed a review of ISI related problems, including welding, BACC and SG ISI, that were identified by the licensee and entered into the corrective action program. The inspectors reviewed the corrective action documents to confirm that the licensee had appropriately described the scope of the problem and had initiated adequate corrective actions. The inspectors performed this review to ensure compliance with 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action requirements. The corrective action documents reviewed by the inspectors are listed in the attachment to this report.

b. Findings

No findings of significance were identified.

1R11 Licensed Operator Requalification

a. Inspection Scope

Annual review of Licensee Requalification Examination Results. On February 16, 2007, the licensee completed the requalification biennial written exam and annual operating tests, required to be given to all licensed operators by 10 CFR 55.59(a)(2). The inspectors performed an in-office review of the overall pass/fail results of the individual written examination and 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 licensees effectiveness in performing routine maintenance activities. This review included an assessment of the licensees practices pertaining to the identification, scope, and handling of degraded equipment conditions, as well as common cause failure evaluations and the resolution of historical equipment problems.

For those systems, structures, and components within the scope of the maintenance rule per 10 CFR 50.65, the inspectors verified that reliability and unavailability were properly monitored, and that 10 CFR 50.65 (a)(1) and (a)(2) classifications were justified in light of the reviewed degraded equipment condition. The inspectors conducted this inspection for two degraded equipment conditions listed below. The inspectors verified that the licensee was appropriately identifying and documenting maintenance rule issues in the corrective action program. The licensees maintenance effectiveness was evaluated for the following two degraded equipment conditions:

• NCR 250154 Reactor coolant system power operated relief valve momentarily opened during control card replacement

• NCR 211171, Diesel breaker would not close during testing

b. Findings

No findings of significance were identified.

1R13 Maintenance Risk Assessments and Emergent Work Control

a. Inspection Scope

The inspectors reviewed the risk impact associated with those activities listed below and verified the licensees associated risk management actions. This review primarily focused on equipment determined to be risk significant within the maintenance rule.

The inspectors also assessed the adequacy of the licensees identification and resolution of problems associated with risk management including emergent work activities. The licensees implementation of compliance procedure CP-253, Power Operation Risk Assessment, was verified in each of the following four work week assessments.

• Work Week 07W39, risk assessment for operation with A train control complex chiller (CHHE-1A) out of service for maintenance, EGDG-1A out of service for testing and emergent work on A train control room emergency ventilation and A channel of the reactor protection system,

• Work Week 07W40, risk assessment for operation with pressurizer power operated relieve valve (PORV) block valve (RCV-11) shut,

• Work Week 07W41, risk assessment for operation with the pressurizer PORV block valve (RCV-11) shut and EGDG-1B and control complex chiller (CHHE-1B)individually out of service for maintenance, and

• Work Week 07W49, risk assessment for operation with EGDG-1A and EGDG-1B individually out of service for surveillance testing and emergent work on reactor protective system A channel.

b. Findings

No findings of significance were identified.

1R15 Operability Evaluations

a. Inspection Scope

The inspectors reviewed the following five NCRs to verify that the operability of systems important to safety was properly established, that the affected components or systems remained capable of performing their intended safety function, and that no unrecognized increase in plant or public risk occurred. The inspectors determined if operability of systems or components important to safety was consistent with technical specifications, the FSAR, 10 CFR Part 50 requirements, and when applicable, NRC Inspection Manual, Part 9900, Technical Guidance, Operability Determinations & Functionality Assessments for Resolution of Degraded or Nonconforming Conditions Adverse to Quality or Safety.

The inspectors reviewed licensee NCRs, work schedules, and engineering documents to check if operability issues were being identified at an appropriate threshold and documented in the corrective action program, consistent with 10 CFR 50, Appendix B requirements; and licensee procedure NGGC-CAP-200, corrective action program.

• NCR 249153, Spurious Trip Of A Reactor Protection Channel

• NCR 250154, Pressurizer PORV momentarily opened

• NCR 251139, Non-Conservative Assumption in Calculation I94-0011 For Reactor Building Service Water Flow Loop

• NCR 255010, Invalid KW and PF Data Recorded During Max Load Test

• NCR 257965, Total Reactivity Worth of Rod Groups 6 and 7 did not meet acceptance criteria

b. Findings

No findings of significance were identified.

1R19 Post-Maintenance Testing

a. Inspection Scope

The inspectors witnessed and/or reviewed post-maintenance test procedures and/or test activities, as appropriate, for selected risk significant systems to verify whether: (1)testing was adequate for the maintenance performed;

(2) acceptance criteria were clear, and adequately demonstrated operational readiness consistent with design and licensing basis documents;

(3) test instrumentation had current calibrations, range, and accuracy consistent with the application;

(4) tests were performed as written with applicable prerequisites satisfied, and

(5) equipment was returned to the status required to perform its safety function. The five post-maintenance tests reviewed are listed below:

• SP-186, AHFL-4A/4B (Control Room) In-Place Filter Testing, and performance testing procedure PT-190, In-Place Filter Testing Procedures, after replacing A train control room charcoal filters per work order (WO) 1130284,

• OP-409, Plant Ventilation System and preventative maintenance procedure PM-136A, Control Complex Chiller CHHE-1A, after performing maintenance on CHHE-1A per WOs 1010776 and 986650,

• SP-354A, Monthly Functional Test of the Emergency Diesel Generator, and Post Maintenance Test WO 01155129, after performing corrective maintenance due to an engine lube oil leak on EGDG-1A,

• SP-630, MUP/HPI Check Valves Full Flow Test (MUP-1A portion only), after performing maintenance on make-up pump MUP-1A per WO 1154513, and

• SP-354B, Monthly Functional Test of the Emergency Diesel Generator EGDG-1B (fast start) after maintenance was performed on EGDG-1B per WOs 803990, 1127159 and 804648.

b. Findings

No findings of significance were identified.

1R20 Refueling and Outage Activities

Refueling Outage (RFO15)

a. Inspection Scope

Outage Planning, Shutdown Monitoring and Licensee Control of Outage Activities The inspectors reviewed the licensees RFO15 Outage Risk Assessment report, to confirm the licensee had appropriately considered risk, industry experience, and previous site-specific problems in developing and implementing the outage plan. During the refueling outage, the inspectors observed and monitored licensee controls over the outage activities listed below. Documents reviewed are listed in the Attachment.

• Outage related risk assessment monitoring

• Controls associated with shutdown cooling, reactivity management, electrical power alignments, containment closure and integrity, and spent fuel pool cooling

• Implementation of equipment clearance activities

• Reduced inventory activities

• Refueling activities

• Reactor mode changes

• Reactor heat-up and pressurization

• Containment cleanup and closeout inspection

• Reactor initial startup and reactor physics testing

• Reactor power ascension and related testing Review of Operating Experience Smart Sample (OPESS) FY2007-03, Crane and Heavy Lift Inspection, Supplemental Guidance for IP-71111.20 The inspectors performed an operating experience smart sample in the area of handling of heavy loads. The inspectors reviewed selected heavy lifting evolutions in the reactor building. Specifically, the inspectors observed removal of the reactor head package to verify the maximum lift height and load path were in accordance with the load drop analysis and reactor vessel head lift procedures. The inspectors verified proper polar crane preventative maintenance was performed prior to the head lift and the crane operators were properly trained and briefed prior to the lift.

b. Findings

Debris found in Containment

Introduction:

A Green, non-cited violation (NCV) of Improved Technical Specification 5.6.1, Procedures, was identified by the NRC for failure to follow procedural guidance associated with removal of debris in containment.

Description:

On December 2, 2007, with the exception of the reactor building incore pit area where limited scope work was continuing, the licensee completed containment inspections in accordance with SP-324, Containment Inspection. The procedure is utilized, in part, to ensure no latent debris is present that can be carried to and possibly block the emergency core cooling system (ECCS) containment sump. The inspection is performed prior to plant heat-up to Mode 4 when the containment sump is required to be operable.

Later on December 2, the inspectors performed a detailed inspection of containment. In general, the inspectors found that the containment was clean and free of debris of substantial size. However, the inspectors identified outage related materials that had not been removed by the licensee. On the 160' elevation, the inspectors found two pairs of work gloves, a bag of tie wraps and two approximately 12" X 18" fibrous scaffold boards under a scaffold storage box. Around reactor coolant pump RCP-1C seal area, the inspectors found a skull cap, 1/2 of a scrubbing pad, pieces of wire and rubber booties.

The debris identified by the inspectors was removed by the licensee prior to Mode 4 entry. As a result of the inspectors observations, the licensee performed additional containment cleanup/inspections to ensure latent debris was removed.

Analysis:

The inspectors determined that the failure to ensure latent debris was removed from containment is a performance deficiency. The finding is more than minor because it could be reasonably viewed as a precursor to a significant event involving debris accumulation on the containment sump screens which could cause impairment to ECCS recirculation flow during a design basis loss of coolant accident. The inspectors referenced Inspection Manual Chapter 0609, Significance Determination Process (SDP),

Phase 1 screening and determined the finding to be of very low safety significance.

Although the debris impacted the mitigating system cornerstone, it would not have resulted in an actual loss of safety function and was not potentially risk significant due to possible external events. A contributing cause of this finding is related to the crosscutting area of Human Performance, specifically Work Practices in that the licensee did not adequately comply with a containment inspection procedure. (IMC 305, H.4(b))

Enforcement:

Improved Technical Specification 5.6.1.1.a requires that written procedures be implemented for those systems referenced in Regulatory Guide 1.33, Revision 2, Appendix A, which would include operation of the emergency core cooling system. SP-324, Containment inspection, Revision 56, enclosures 5, 8 and 9 require, in part, that no latent debris exist in containment that could impact performances of the ECCS. Contrary to this, on December 2, 2007, the licencee documented satisfactory completion of the SP-324 enclosures, even though the acceptance criteria regarding debris was not met. Because this finding is of very low safety significance and because it was entered into the licensees corrective action program (NCR 257426), this violation is being treated as an Non-cited Violation consistent with Section VI.A of the NRC Enforcement Policy, and is identified as NCV 05000302/2007005-001, Failure to Follow Procedural Guidance Associated with Removal of Containment Debris. Corrective actions completed include: removal of the debris identified by the inspectors and performing additional inspection and cleaning of containment.

1R22 Surveillance Testing

a. Inspection Scope

The inspectors observed and/or reviewed the surveillance tests listed below to verify that technical specification surveillance requirements were followed and that test acceptance criteria were properly specified. The inspectors verified that proper test conditions were established as specified in the procedures, that no equipment preconditioning activities occurred, and that acceptance criteria had been met. Additionally, the inspectors also verified that equipment was properly returned to service and that proper testing was specified and conducted to ensure that the equipment could perform its intended safety function following maintenance or as part of surveillance testing. The following seven activities were observed/reviewed:

In-Service Test:

• SP-344A, RWP-2A, SWP-1A and Valve Surveillance Surveillance Tests:

• SP-354A, Monthly Functional Test Of The Emergency Diesel Generator EGDG-1A

• SP-349C, EFP-3 and Valve Surveillance (Cold Shutdown Valve Testing)

• SP-102, Control Rod Drop Time Tests

• SP-417, Refueling Interval Integrated Plant Response to an Engineered Safeguards Actuation Containment Isolation Valve Test:

• SP-179C, Containment Leakage Test - Type C, make-up valve MUV-253

• SP-179C, Containment Leakage Test - Type C, leak rate valve LRV-50

b. Findings

No findings of significance were identified.

RADIATION SAFETY

Cornerstone: Occupational Radiation Safety (OS)

2OS1 Access Controls to Radiologically Significant Areas

a. Inspection Scope

Access Controls The inspectors evaluated licensee guidance and its implementation for controlling worker access to radiologically significant areas and monitoring jobs in-progress. The inspectors directly observed implementation of administrative and physical radiological controls; evaluated radiation worker (radworker) and health physics technician (HPT) knowledge of and proficiency in implementing radiation protection requirements; and assessed worker exposures to radiation and radioactive material.

During facility tours, the inspectors directly observed postings and physical controls for radiation areas and high radiation areas (HRAs) established within the radiologically controlled area (RCA) of the auxiliary building and radioactive waste (radwaste)processing and storage locations. The inspectors independently measured radiation dose rates or directly observed conduct of licensee radiation surveys for selected RCA areas. Results were compared to current licensee surveys and assessed against established postings and Radiation Work Permit (RWP) controls. Licensee key control and access barrier effectiveness were observed and evaluated for selected Locked High Radiation Area (LHRA) locations. Changes to procedural guidance for LHRA and Very High Radiation Area controls were discussed in detail with health physics (HP)supervisors. Physical controls for storage of irradiated material within the spent fuel pool were observed. In addition, licensee controls for areas where dose rates could change significantly as a result of refueling operations or radwaste activities were reviewed and discussed.

The inspectors attended the pre-job briefings and remotely observed activities leading to the removal of the core barrel as part of the 10 year in-service inspection. The observations afforded the inspectors opportunity to assess various aspects of the radiation protection program including communications between HP and various workgroups, internal communications, RWP controls, contamination control, surveys, radiation worker adherence to RWP and other HP guidance, HPT proficiency in providing job coverage and supervisory willingness to intervene when conditions deviated from expected.

The inspectors observed various other work activities including work on pressurizer surge line as part of the alloy 600 inspection and weld overlay. The inspectors continued to evaluate various aspects of the HP program to include comparisons of the electronic dosimeter (ED) alarms setpoints with area radiation survey results and ED alarm response actions were discussed with radiation workers and HP supervisors.

The inspectors evaluated the effectiveness of radiation exposure controls, including air sampling, barrier integrity, engineering controls, and postings through a review of both internal and external exposure results. Licensee evaluations of skin dose resulting from discrete radioactive particle or dispersed skin contamination events were reviewed and assessed. For HRA tasks involving significant dose rate gradients, the inspectors evaluated procedural guidance for the use and placement of whole body and extremity dosimetry to monitor worker exposure. The inspectors also reviewed and discussed internal dose assessments and whole body count results for one individual.

Radiation protection activities were evaluated against the requirements of Final Safety Analysis Report (FSAR) Chapter 11; Technical Specifications (TS) Section 5.8; 10 Code of Federal Regulations (CFR) Part 20; and approved licensee procedures. Records reviewed are listed in Section 2OS1 of the report Attachment.

Problem Identification and Resolution Licensee Corrective Action Program (CAP)documents associated with access control to radiologically significant areas were reviewed and assessed. This included the review of a licensee self-assessment and selected Nuclear Condition Reports (NCRs) related to radworker and HPT performance.

The inspectors evaluated the licensees ability to identify, characterize, prioritize, and resolve the identified issues in accordance with CAP-NGGC-0200, Corrective Action Program, Rev. 19. Licensee CAP documents reviewed are listed in Section 2OS1 of the report Attachment.

The inspectors completed 21 of the specified line-item samples detailed in Inspection Procedure (IP) 71121.01.

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 refueling outage 15 (R15) 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:

Alloy 600 weld overlay work, steam generator (S/G) eddy current testing, and scaffolding emplacement and removal. 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 an ALARA review committee meeting 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 S/G eddy current testing and Alloy 600 work. 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 S/G eddy current testing, Alloy 600 remediation, and reactor building scaffold 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 reactor building 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.

Plant exposure history for calendar year (CY) 2004 through CY 2006 and data reported to the NRC pursuant to 10 CFR 20.2206 were reviewed, as were established goals for reducing collective exposure during the current R15 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.

ALARA program activities and their implementation were reviewed against 10 CFR Part 20, and approved licensee 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. Procedures and records reviewed within this inspection area are listed in Sections 2OS2 of the report Attachment.

Problem Identification and Resolution The inspectors reviewed selected NCRs in the area of exposure control. The inspectors evaluated the licensees ability to identify, characterize, prioritize, and resolve the identified issues in accordance with CAP-NGGC-0200, Corrective Action Program, Rev. 19. The inspectors also evaluated the scope of the licensees internal audit program and reviewed recent assessment results.

Documents reviewed for problem identification and resolution are listed in Section 2PS2 of the report Attachment.

The inspectors completed 18 of the specified line-item samples detailed in IP 71121.02.

b. Findings

No findings of significance were identified.

Cornerstone: Public Radiation Safety (PS)

2PS1 Radioactive Gaseous and Liquid Effluent Treatment and Monitoring Systems

a. Inspection Scope

Current licensee programs for monitoring, tracking, and documenting the results of both routine and abnormal liquid releases to onsite and offsite surface and ground water environs were reviewed and discussed in detail. The inspectors discussed results of the recently completed site hydrology study which confirmed groundwater movement toward the west. Radionuclide analysis results for samples collected from eight shallow wells and one deep well surrounding the power block, and for four shallow wells surrounding a leaching pond which receives intermittent releases containing low concentrations of tritium from the turbine building and is located within the owner controlled area were discussed in detail. Low concentrations of tritium, less than or equal to 1000 picocuries per liter (pCi/l) were reported from wells located west of the power block and in wells adjacent to the onsite leaching pond. All values were significantly less than the Environmental Protection Agency drinking water limit of 20,000 pCi/l. No other nuclides have been detected in the samples collected. In addition, the inspectors reviewed and discussed the status of 10 CFR 50.75

(g) spill data. Licensee current capabilities and proposed surveillances to minimize and rapidly identify any abnormal leaks from onsite systems, structures, and components were reviewed and discussed. In addition, the inspectors reviewed and discussed current licensee guidance for reporting any potential releases to offsite groundwater environs.

The inspectors completed two of the specified radiation protection line-item samples detailed in IP 71122.01.

b. Findings

No findings of significance were identified.

2PS2 Radioactive Material Processing and Transportation

a. Inspection Scope

Waste Processing and Characterization During inspector walk-downs, accessible sections of the liquid radwaste processing system were assessed for material condition and conformance with system design diagrams. Inspected equipment included liquid waste demineralizer skids and abandoned radwaste processing equipment. In addition, the inspectors discussed recent processing system engineering changes, and radwaste program implementation with licensee staff.

The 2006 Radioactive Effluent Report and radionuclide characterizations from 2005 -

2007 for selected waste streams were reviewed and discussed with radwaste staff. For Primary Resin and Dry Active Waste (DAW) the inspectors evaluated analyses for hard-to-detect nuclides, reviewed the use of scaling factors, and examined comparison results between licensee waste stream characterizations and outside laboratory data.

The inspectors also reviewed the licensees procedural guidance for monitoring changes in waste stream isotopic mixtures.

Radwaste processing activities and equipment configuration were reviewed for compliance with the licensees Process Control Program (PCP) and FSAR, Chapter 11.

Waste stream characterization analyses were reviewed against regulations detailed in 10 CFR Part 20, 10 CFR Part 61, and guidance provided in the Branch Technical Position on Waste Classification and Waste Form. Reviewed documents are listed in Section 2PS2 of the report Attachment.

Transportation The inspectors directly observed preparation activities for a shipment of contaminated laundry. The inspectors noted package markings and placarding, performed independent dose rate measurements, and interviewed shipping technicians regarding Department of Transportation (DOT) regulations.

The inspectors directly observed and evaluated staff proficiency in preparation of radioactive material shipment number 07-075. In addition, shipping records for four previous radioactive material or radwaste shipments were reviewed for consistency with licensee procedures and compliance with NRC and DOT regulations. The inspectors reviewed emergency response information, DOT shipping package classification, radiation survey results, and evaluated whether receiving licensees were authorized to accept the packages. For selected shipment records, the licensees handling of Type B shipping casks was compared to Certificate of Compliance (CoC) requirements. In addition, training records and training curricula for individuals currently qualified to prepare shipments of radioactive material were reviewed.

Transportation program implementation was reviewed against regulations detailed in 10 CFR Part 20, 10 CFR Part 71, 49 CFR Parts 172-178; as well as the guidance provided in NUREG-1608. Training activities were assessed against 49 CFR Part 172 Subpart H. Documents reviewed during the inspection are listed in Section 2PS2 of the report Attachment.

Problem Identification and Resolution Selected NCRs in the area of radwaste/shipping were reviewed in detail and discussed with licensee personnel. The inspectors assessed the licensees ability to characterize, prioritize, and resolve the identified issues in accordance with licensee procedure CAP-NGGC-0200, Corrective Action Program, Rev. 19. The inspectors also evaluated the scope of the licensees internal audit program and reviewed recent assessment results. Documents reviewed for problem identification and resolution are listed in Section 2PS2 of the report Attachment.

The inspectors completed six of the required samples specified in IP 71122.02.

b. Findings

No findings of significance were identified.

4OA1 Performance Indicator (PI) Verification

.1 Reactor Safety Performance Indicators

a. Inspection Scope

The inspectors checked the accuracy of the mitigating systems performance indicators listed below to verify the accuracy of the PI data reported. Performance indicator data submitted from October 2006 through September 2007 was compared for consistency to data obtained through the review of, monthly operating reports, nuclear condition reports, and control room records. The inspections were conducted in accordance with NRC Inspection Procedure 71151, Performance Indicator Verification. The applicable planning standard, Nuclear Energy Institute (NEI) 99-02, Revision 5, Regulatory Assessment Performance Indicator Guidelines, and the licensees procedure P06-0002, NRC Mitigating System Performance Index (MSPI) Basis Document For The CR3 Nuclear Plant, were used to check the reporting for each data element. The inspectors discussed the PI data with licensee personnel associated with performance indicator data collection and evaluation.

• Emergency AC power

• Residual Heat Removal system

• Heat Removal System

• High Pressure Injection System

• Cooling Water System

b. Findings

No findings of significance were identified.

.2 Radiation Protection Performance Indicators

a. Inspection Scope

Occupational Radiation Safety Cornerstone The inspectors reviewed the Occupational Exposure Control Effectiveness PI results for the Occupational Radiation Safety Cornerstone from December 1 to November 24, 2007 For the assessment period, the inspectors reviewed ED alarm logs, monthly PI reports, and selected NCRs 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 sections 2OS1 and 4OA1 of the report Attachment.

Public Radiation Safety Cornerstone The inspectors reviewed the Radiological Control Effluent Release Occurrences PI results for the Public Radiation Safety Cornerstone from January through September, 2007. For the assessment period, the inspectors reviewed cumulative and projected doses to the public, out-of-service effluent radiation monitors and compensatory sampling data, and selected NCRs related to Radiological Effluent Technical Specifications/Offsite Dose Calculation Manual issues. The inspectors also reviewed licensee procedural guidance for collecting and documenting PI data.

Documents reviewed are listed in section 4OA1 of the report Attachment.

The inspectors completed two of the required samples specified in IP 71151.

b. Findings

No findings of significance were identified.

4OA2 Problem Identification and Resolution

.1 Daily Screening of Items Entered Into the Corrective Action Program

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 (CAP). This review was accomplished by attending daily plant status meetings, interviewing plant operators and applicable system engineers, and accessing the licensees computerized database.

b. Findings

No findings of significance were identified.

.2 Annual Sample Review

a. Inspection Scope

The inspectors selected NCRs 244022 for a detailed review. The NCR was initiated to address reoccurring increasing particulate levels in onsite diesel fuel tanks. The inspectors checked that the issue had been completely and accurately identified in the licensees CAP, and that safety concerns were properly classified and prioritized for resolution, apparent cause determinations were sufficiently thorough, and appropriate corrective actions were implemented in a manner consistent with safety and compliance with plant technical specifications and 10 CFR 50. The inspectors also evaluated the NCR using the requirements of the licensees CAP as delineated in corrective action procedure CAP-NGGC-200, Corrective Action Program.

b. Findings and Observations

NCR 244022 documented the need to investigate the cause of increasing particulate levels in both safety-related and non-safety-related diesel fuel tanks. The particulate level did not impact the operability of any safety related system. The inspectors determined that the short term corrective action of filtering the tanks was appropriate.

The investigation, utilizing both outside resources and industry operating experience, determined that the cause of the increasing particulate levels was due to the interaction between high and low sulfur diesel fuels when high sulfur fuel was added to the tanks over the last year. The inspectors determined that long-term corrective actions in place should resolve the issue.

.3 Annual Sample Review

a. Inspection Scope

The inspectors performed an in-depth review of eleven nuclear condition reports (NCR)that described issues associated with plant status control (PSC), including a priority one NCR that identified a significant adverse trend in PSC. The inspectors reviewed the licensees self assessment report on this subject, the corrective actions that were implemented to address the overall programmatic PSC issues, and the investigative methods and time-lines the licensee will use in investigating and correcting new plant status issues. During the inspection period, the inspectors observed meetings held by the newly formed Plant Status Control Review Board, that performed an independent review of completed NCR investigations and the respective corrective actions associated with plant status control issues. Additionally, the inspectors reviewed the newly formed Plant Status Control Event Zero-Tolerance Policy that was implemented during the inspection period.

b. Findings and Observations

No findings of significance were identified. The inspectors found that the licensee has been very aggressive in their investigations to understand and to correct the causes of plant status control (PSC) issues. Corrective actions include increasing awareness on plant status control on a plant wide level, i.e., PSC cultural change, provide better guidance on PSC issue investigations, and upper management involvement in reviewing completed investigations. For example, the licensee has developed a procedure that provides better guidance for interviewing personnel and documenting facts relating to PSC issues, and guidance on the timeliness to complete PSC issue investigations are now provided. With the creation of the PSC Review Board which consisted of members from upper management (PGM, Operations manager, Maintenance manager, Chemistry superintendent, and Human Performance team lead) and the Plant Status Control Zero-Tolerance Policy, upper management is directly involved in reviewing all PSC issue investigations and corrective actions. Training, PSC operating experience discussions at pre-job briefs, and more personnel accountability are other methods the licensee is using to heighten the awareness of the importance to ensure good plant status control.

.4 Semi-Annual Trend Review

a. Inspection Scope

As required by Inspection Procedure 71152, Identification and Resolution of Problems, the inspectors performed a review of the licensees CAP and associated documents to identify trends that could indicate the existence of a more significant safety issue. The inspectors review was focused on repetitive equipment issues, but also considered the results of daily inspector CAP item screening discussed in section 4OA2.1 above, plant status reviews, plant tours, and licensee trending efforts. The inspectors review nominally considered the six month period of June 2007 through December 2007. The review also included issues documented in the Equipment Performance Priority List dated December, 2007; System Health Report January to June 2007 dated September 13, 2007; various nuclear assessment section reports and maintenance rule assessments. The inspectors compared and contrasted their results with the results contained in the licensees 3rd Quarter 2007, Site CAP Rollup & Trend Analysis report.

Corrective actions associated with a sample of the issues identified in the licensees trend report were reviewed for adequacy.

b. Assessment and Observations No findings of significance were identified. The inspectors evaluated the licensees trend methodology and observed that the licensee had performed a detailed review. The inspectors compared the licencees reviews with the results of the inspectors daily screening and did not identify any discrepancies or potential trends in the data which the licensee had failed to identify.

40A3 Followup of Events and Notices of Enforcement Discretion Operator performance during non-routine event

a. Inspection Scope

For the four non-routine plant evolutions described below, the inspectors reviewed the operating crews performance, operator logs, control board indications, and the plant computer data to verify that operator response was in accordance with plant procedures.

• October 10, Pressurizer power operated relief valve opened following card replacement,

• October 29, Rapid power decrease to 65 percent power in accordance with licensee abnormal procedure AP-510, Rapid Power Reduction, following a loss of the condensate pump CDP-1B,

• November 3, Reactor shutdown to Mode 3 in accordance with SP-209A, Plant Shutdown And Cooldown, and

• December 7, Plant startup activities in accordance with OP-203, Plant Startup.

b. Findings

No findings of significance were identified.

4OA5 Other Activities

(Closed) Temporary Instruction (TI) 2515/166, Pressurized Water Reactor Containment Sump Blockage (NRC Generic Letter 2004-02)

a. Inspection Scope

The inspectors reviewed Unit 3 implementation of the licensee's commitments documented in their September 1, 2005 response to Generic Letter 2004-02, Potential Impact of Debris Blockage on Emergency Recirculation During Design Basis Accidents at Pressurized Water Reactors. These commitments included the permanent modification of the Containment Building ECCS sump strainer assembly, and installation of a flow distributor and debris interceptor. The inspectors reviewed the sump strainer assembly Engineering Change packages (EC), corresponding 10 CFR 50.59 evaluation, and ECCS sump inspection requirements in the Plant Operating Manual. The inspectors conducted a visual walkdown to verify the installed strainer assembly configuration was consistent with drawings and specifications provided in the ECs.

b. Findings and Observations

No findings of significance were identified.

The inspectors determined the following answers to the Reporting Requirements detailed in TI 2515/166-05 issued 5/16/07:

05.a Progress Energy implemented plant modifications and procedure changes at Crystal River committed to in their GL 2004-02 response for Unit 3. A list of commitments and their respective completion dates is listed in the attachment, Status of GL 2004-02 Commitments for Crystal River 3.

05.b Progress Energy updated the Crystal River 3 licensing bases to reflect the corrective actions taken in response to GL 2004-02.

05.c By NRC letter (TAC NO. MC4678) an extension has been approved to extend the completion of steam generator fibrous insulation removal to the end of the Fall 2009 refueling outage during which the steam generators will be replaced. An extension was also approved by the same letter for completion of the in-vessel downstream effects and ex-vessel downstream effects evaluations by February 29, 2008.

TI 2515/166 is closed for Crystal Unit 3, no additional modifications or procedural changes under GL 2004-02 are anticipated.

4OA6 Meetings

Exit Meeting Summary

On January 7, 2008, the resident inspectors presented the inspection results to Mr. D.

Young, Site Vice President and other members of licensee management, who acknowledged the finding. The inspectors confirmed that proprietary information was not provided or examined during the inspection.

ATTACHMENT:

SUPPLEMENTAL INFORMATION

KEY POINTS OF CONTACT

Licensee

M. Annacone, Plant General Manager

W. Brewer, Manager, Maintenance

S. Cahill, Manager, Engineering

P. Dixon, Manager, Nuclear Assessment

J. Franke, Director of Site Operations

R. Hons, Manager, Training

J. Holt, Manager, Operations

D. Herrin (Acting), Supervisor, Licensing

M. Rigsby, Superintendent, Radiation Protection

J. Stephenson, Supervisor, Emergency Preparedness

Ivan Wilson, Manager, Outage and Scheduling

D. Young, Vice President, Crystal River Nuclear Plant

NRC personnel

S. Vias, Chief, Reactor Projects Branch 3, NRC Region II

LIST OF ITEMS

OPENED AND CLOSED

Opened and Closed

NCV

05000302/2007005-001 NCV Failure to Follow Procedural Guidance Associated with Removal of Containment Debris (Section 1R20)

Closed

Temporary Instruction (TI) 2515/166, Pressurized Water Reactor Containment Sump Blockage (NRC Generic Letter 2004-02) (Section 4OA5)