IR 05000382/2004006

IR 05000382-04-006; 05/03 - 21/2004; Waterford Steam Electric Station, Unit 3. Identification and Resolution of Problems, Mitigating Systems, Barrier Integrity

ML041890343
Person / Time
Site:	Waterford
Issue date:	07/06/2004
From:	Laura Smith Division of Reactor Safety IV
To:	Venable J Entergy Operations
References
IR-04-006
Download: ML041890343 (29)

Text

uly 6, 2004

SUBJECT:

WATERFORD STEAM ELECTRIC STATION, UNIT 3 - NRC PROBLEM IDENTIFICATION AND RESOLUTION INSPECTION REPORT 05000382/2004006

Dear Mr. Venable:

On May 21, 2004, the NRC completed an inspection at your Waterford Steam Electric Station, Unit 3. The enclosed report documents the inspection findings, which were discussed on May 21, 2004, with you and other members of your staff.

This inspection was an examination of activities conducted under your license as they relate to the identification and resolution of problems, and compliance with the Commissions rules and regulations and with the conditions of your license. Within these areas, the inspection consisted of selected examination of procedures and representative records, observations of activities, and interviews with personnel.

The team reviewed approximately 135 corrective action program documents, apparent and root cause analyses and plant procedures for the identification and resolution of problems. Based on this review, the team found that your processes to identify, prioritize, evaluate, and correct problems were generally effective; thresholds for identifying issues remained appropriately low and, in most cases, corrective actions were adequate to address conditions adverse to quality.

However, a number of issues were identified associated with the proper identification, evaluation and correction of degraded conditions in the plant. Most of these issues were identified when the team reviewed corrective actions associated with longstanding degraded conditions and design issues at Waterford 3, which had cross-cutting aspects in the area of problem identification and resolution.

This report documents three findings that were evaluated under the risk significance determination process as having very low safety significance (Green). The NRC has also determined that violations were associated with these findings. The violations are being treated as noncited violations consistent with Section VI.A of the Enforcement Policy. If you contest the violations or significance of these noncited violations, you should provide a response within 30 days of the date of this inspection report, with the basis for your denial, to the U.S. Nuclear Regulatory Commission, ATTN: Document Control Desk, Washington, DC 20555-0001, with

Entergy Operations Inc. -2-copies to the Regional Administrator, U.S. Nuclear Regulatory Commission, Region IV, 611 Ryan Plaza Drive, Suite 400, Arlington, Texas 76011; the Director, Office of Enforcement, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001; and the NRC Resident Inspector at the Waterford Steam Electric Station, Unit 3 facility.

In accordance with 10 CFR 2.390 of the NRCs "Rules of Practice," a copy of this letter, its enclosure, and your response will be made available electronically for public inspection in the NRC Public Document Room or from the Publicly Available Records component of NRCs document system (ADAMS). ADAMS is accessible from the NRC Web site at http://www.nrc.gov/reading-rm/adams.html (the Public Electronic Reading Room).

Sincerely,

//RA//

Linda Joy Smith, Chief Plant Engineering Branch Division of Reactor Safety Docket: 50-382 License: NPF-38 Enclosure:

NRC Inspection Report 05000382/2004006 W/attachment: Supplemental Information copy to:

Senior Vice President and Chief Operating Officer Entergy Operations, Inc.

P.O. Box 31995 Jackson, MS 39286-1995 Vice President, Operations Support Entergy Operations, Inc.

P.O. Box 31995 Jackson, MS 39286-1995 Wise, Carter, Child & Caraway P.O. Box 651 Jackson, MS 39205

Entergy Operations Inc. -3-General Manager, Plant Operations Waterford 3 SES Entergy Operations, Inc.

17265 River Road Killona, LA 70066-0751 Manager - Licensing Manager Waterford 3 SES Entergy Operations, Inc.

17265 River Road Killona, LA 70066-0751 Chairman Louisiana Public Service Commission P.O. Box 91154 Baton Rouge, LA 70821-9154 Director, Nuclear Safety &

Regulatory Affairs Waterford 3 SES Entergy Operations, Inc.

17265 River Road Killona, LA 70066-0751 Michael E. Henry, State Liaison Officer Department of Environmental Quality Permits Division P.O. Box 4313 Baton Rouge, LA 70821-4313 Parish President St. Charles Parish P.O. Box 302 Hahnville, LA 70057 Winston & Strawn 1400 L Street, N.W.

Washington, DC 20005-3502

Entergy Operations Inc. -4-Electronic distribution by RIV:

Regional Administrator (BSM1)

DRP Director (ATH)

DRS Director (DDC)

Senior Resident Inspector (MCH)

Branch Chief, DRP/E (WBJ)

Senior Project Engineer, DRP/E (VGG)

Staff Chief, DRP/TSS (PHH)

RITS Coordinator (KEG)

Jennifer Dixon-Herrity, OEDO RIV Coordinator (JLD)

WAT Site Secretary (AHY)

ADAMS: / Yes * No Initials: __ljs____

/ Publicly Available * Non-Publicly Available * Sensitive / Non-Sensitive DOCUMENT: R\_WAT\2004\WT2004-06RP-PJA.wpd RIV:DRS\C:PEB DRP/B:SRI PEB:RI DRP/E:SRI DRP/E:SRI LJSmith:jlh PJAlter PAGoldberg GDReplogle MCHay RLNease for via e-mail via e-mail via e-mail via e-mail 07/02 /04 7/06/04 07/01/04 07/02/04 07/02/04 DRP/E:RI C:DRP/E C:PEB GFLarkin WDJones LJSmith via e-mail /RA/ /RA/

07/02/04 07/02 /04 07/06 /04 OFFICIAL RECORD COPY T=Telephone E=E-mail F=Fax

ENCLOSURE U.S. NUCLEAR REGULATORY COMMISSION

REGION IV

Docket: 50-382 License: NPF-38 Report: 05000382/2004006 Licensee: Entergy Operations, Inc.

Facility: Waterford Steam Electric Station, Unit 3 Location: Hwy. 18 Killona, Louisiana Dates: May 3 through May 21, 2004 Inspectors: P. Alter, Senior Resident Inspector, Projects Branch B P. Goldberg, Reactor Inspector, Plant Engineering Branch G. Replogle, Senior Resident Inspector, Projects Branch E M. Hay, Senior Resident Inspector, Projects Branch E G. Larkin, Resident Inspector, Projects Branch E Approved by: L. J. Smith, Chief Plant Engineering Branch Division of Reactor Safety Enclosure

SUMMARY OF FINDINGS

IR 05000382/2004006; 05/03 - 21/2004 ; Waterford Steam Electric Station, Unit 3.

Identification and Resolution of Problems, Mitigating Systems, Barrier Integrity The inspection was conducted by three senior resident inspectors, one reactor inspector, and one resident inspector. Three green findings of very low safety significance were identified during the inspection and were classified as noncited violations. The significance of most findings is indicated by their color (green, white, yellow, red) using IMC 0609, Significance Determination Process. Findings for which the significant determination process does not apply may be green or assigned a severity level after NRC management review. The NRCs program for overseeing the safe operation of commercial nuclear power reactors is described in NUREG-1649, Reactor Oversight Process, Revision 3, dated July 2000.

Identification and Resolution of Problems

• The team reviewed approximately 135 corrective action program documents, apparent and root cause analyses and plant procedures for the identification and resolution of problems. Based on this review, the team found that the licensees process to identify, prioritize, evaluate, and correct problems was generally effective; thresholds for identifying issues remained appropriately low and, in most cases, corrective actions were adequate to address conditions adverse to quality. However, a number of issues were identified associated with the proper identification, evaluation and correction of degraded conditions in the plant. Most of these issues were identified when the team reviewed corrective actions associated with longstanding degraded conditions and design issues at Waterford 3, which had cross-cutting aspects in the area of problem identification and resolution.

The team concluded that a positive safety-conscience work environment exists at Waterford 3. The team determined that employees and contractors feel free to raise safety concerns to their supervision or bring concerns to the employees concern program.

A. Inspector-Identified and Self-Revealing Findings

Cornerstone: Mitigating Systems

Green.

The team identified a 10 CFR 50, Appendix B, Criterion XVI, noncited violation for situations where the licensee failed to promptly correct conditions adverse to quality associated with the main feed isolation valve hydraulic actuating systems. In two cases, the licensee failed to promptly correct instances where the hydraulic actuator thermal relief valves failed to properly function. Consequently, the hydraulic portion of the valve actuator experienced repetitive over-pressure conditions. In one case, the licensee failed to properly address system operability and, for a two-week period, actual valve operability was unknown.

The issue was more than minor because it affected the mitigating systems cornerstone objective to ensure the availability of systems that respond to initiating events. The finding was determined to be of very low risk significance because each issue: was not a design or qualification deficiency; did not result in the loss of a safety system; did not represent an actual loss of a safety function of a single train for greater than its technical specification allowed outage time; did not represent an actual loss of safety function of one or more non-Technical Specification trains of equipment designated as risk significant per 10 CFR 50.65 for greater than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />; and was not potentially risk significant due to a seismic, fire, flooding, or severe weather initiating event. Because the failure to promptly identify and correct the over-pressure condition was of very low safety significance and has been entered into the licensees corrective action program as condition reports CR-WF3-2004-1533, 1540 and 1551, this violation is being treated as a noncited violation, consistent with Section VI.A of the NRC Enforcement Policy (Section 4OA2 e.).

Cornerstone: Barrier Integrity

Green.

The team identified a noncited violation of 10 CFR 50, Appendix B, Criterion XVI, for the failure to promptly identify and correct a condition adverse to quality.

Specifically, on multiple occasions the licensee failed to identify and correct an inappropriate value of the unfiltered inleakage parameter used to calculate the control room operator dose for design basis accident conditions involving radiological releases.

This failure resulted in significantly underestimating the actual dose to operators.

This finding was greater than minor because it affected the barrier integrity cornerstone objective related to design control of the control room envelope and was determined to be of very low safety significance because the deficiency only affected the radiological barrier function provided for the control room. Because the failure to promptly identify and correct the analysis was of very low safety significance and has been entered into the licensees corrective action program as Condition Report CR-WF3-2004-1403, this violation is being treated as a noncited violation, consistent with Section VI.A of the NRC Enforcement Policy (Section 4OA2 e.).

Green.

The team identified a noncited violation of 10 CFR 50, Appendix B, Criterion XVI, for the failure to promptly identify and correct a condition adverse to quality.

Specifically, on multiple occasions the licensee failed to correct a known deficient condition involving the failure to account for instrument uncertainty to satisfy Technical Specification Surveillance Requirement 4.7.6.5.a. This failure potentially affects the ability of the control room envelope to perform its design function with respect to protecting operators from postulated design basis accidents resulting in radiological releases.

This finding was greater than minor because it affected the barrier integrity cornerstone objective related to maintaining the barrier function of the control room envelope. The finding was determined to be of very low safety significance because the deficiency only affected the radiological barrier function provided for the control room. Because the failure to promptly identify and correct the analysis was of very low safety significance and has been entered into the licensees corrective action program as condition report CR-WF3-2004-1561, this violation is being treated as a noncited violation, consistent with Section VI.A of the NRC Enforcement Policy (Section 4OA2 e.).

Licensee-Identified Violations

None.

REPORT DETAILS

4OA2 Identification and Resolution of Problems

a. Effectiveness of Problem Identification

(1) Inspection Scope The inspectors reviewed items selected across the seven cornerstones to determine if problems were being properly identified, characterized, and entered into the corrective action program for evaluation and resolution. Specifically, the teams review included a selection of approximately 135 condition reports. The majority were opened or closed since the last NRC Problem Identification and Resolution Inspection completed on December 20, 2002. The team also performed a historical review of condition reports written over the last five years for the high pressure safety injection system, emergency feedwater system, safety-related battery chargers and the emergency diesel generators.

The team reviewed a sample of licensee audits and self assessments, trending reports, system health reports, and various other reports and documents related to the problem identification and resolution program. The audit and self-assessment results were compared with the self-revealing and NRC-identified issues to determine the effectiveness of the audits and self assessments.

The team interviewed station personnel and evaluated corrective action documentation to determine the licensees threshold for identifying problems and entering them into the corrective action program. The team attended morning work planning and work request classification meetings to evaluate the licensees evaluation of plant issues against corrective action program criteria for condition report initiation. The team evaluated the licensees efforts in establishing the scope of problems by reviewing control room operator logs, security and radiation protection logs and work orders (formerly maintenance action items).

In addition, the team reviewed the licensees evaluation of selected industry experience information, including operator event reports and NRC and generic vendor notices, to assess if issues applicable to Waterford 3 were appropriately addressed.

A listing of specific documents reviewed during the inspection is included in the attachment to this report.

(2) Assessment The team determined that, in general, problems were adequately identified and entered into the corrective action program. The threshold for entering issues into the corrective action program was appropriately low. Conditions adverse to quality identified in other systems, such as the work management system and various logs were properly entered into the licensees corrective action program.

However, the team found one example of ineffective problem identification during this inspection. One example of ineffective problem identification was identified during a previous inspection. These conditions related to a long-standing design issue for Waterford 3.

Example 1 - Noncited Violation 05000382/2004006-01: Failure to Promptly Identify Inappropriate Assumption and Correct Control Room Operator Dose Analysis The team determined problem identification was not adequate based on the number of opportunities the licensee had to identify the inappropriate assumptions beginning from the design change that was made in 1983. The licensee had additional opportunities to identify this design deficiency when revisions to the dose calculation were performed (1994 and 1998) that actually included changes to the assumed leakage values. In March 2004, the licensee identified that the dose analysis for operators in the control room during a radiological emergency assumed that the control room was in the pressurized mode contrary to design basis documents. (This issue is discussed in more detail in Report Section 4OA2 e.)

Example 2 - Noncited Violation 05000382/2003006-04: Inadequate Design Control of Switchgear Ventilation System.

NRC inspection report 05000382/2003006 documented a design control violation with cross-cutting aspects related to inadequate problem identification. The licensee missed two prior opportunities to identify the need to correct their design for the of the switchgear ventilation system. The February 2, 1989 problem evaluation and information request Number 10672, failed to determine the basis for the safety-related position of inlet Dampers SVS-101 and 102. The 1997 design basis reconstitution process Item SVS-01-058 did not determine the basis for the normal and safety-related position of the inlet dampers. Yet Final Safety Analysis Report Section 6.4.2.4 states, in part, that the ventilation zones adjacent to the control room are to be maintained always negative with respect to the [control room] envelope. The switchgear ventilation zone is adjacent to the main control room.

b. Prioritization and Evaluation of Issues

(1) Inspection Scope The team reviewed condition reports, engineering operability evaluations and operations operability determinations to assess the licensees ability to evaluate the importance of the conditions adverse to quality. The team reviewed the results of corrective action review group meetings that assigned significance and priority to condition reports. The team reviewed a sample of failure mode analyses, apparent cause analyses and root cause analyses, to ascertain whether the licensee identified and considered the full extent of conditions, generic implications, common causes, and previous occurrences.

The team also observed management oversight of the significant conditions adverse to quality including one Corrective Action Review Board meeting.

In addition, the inspectors reviewed licensee evaluations of selected industry operating experience information, including operating event reports and NRC and generic vendor notices, to assess whether issues applicable to Waterford 3 were appropriately addressed. The team performed a historical review of condition reports covering the last five years regarding the high pressure safety injection system, the emergency feedwater system, safety-related battery chargers and the emergency diesel generators to determine if the licensee had appropriately addressed long-standing issues and those that might be age dependent.

A listing of specific documents reviewed during the inspection is included in the attachment to this report.

(2) Assessment The team concluded that problems were generally prioritized and evaluated in accordance with the licensees corrective action program guidance and NRC requirements. The team found that for the sample of root cause analyses reviewed, that the licensee was generally self critical and exhaustive in its research onto the history of significant conditions adverse to quality.

However, the team found two examples of ineffective problem evaluation during this inspection. One example of ineffective problem evaluation was identified during a previous inspection. In two cases, the condition related to a long-standing design issue for Waterford 3.

Example 1 - Noncited Violation 05000382/2004006-02: Failure to Promptly Correct Over-pressure Condition in Main Feed Water Isolation Valve Hydraulic Operating Systems.

The team determined that the licensee failed to effectively evaluate operability of a main feed water isolation valve following the February 20, 2004 hydraulic operating system over-pressure condition. The licensee stated that accumulator pressure was reduced therefore Valve FW-184B was operable. However, the operability assessment did not address the potential for the system to over-pressurize again. As a result, the licensee did not start additional monitoring of hydraulic pressure until March 4, 2004 and then only on an infrequent basis. (This issue is discussed in more detail in Report Section 4OA2 e.)

Example 2 - Noncited Violation 05000382/2004006-03: Failure to Promptly Identify and Correct a Known Deficient Condition Involving the Failure to Account for Instrument Uncertainty to Satisfy Technical Specification Surveillance Requirement 4.7.6.5.a.

The team determined that the licensee failed to effectively evaluate regulatory requirements. On two occasions this issue was entered into the corrective action process, with two different reasons as to why the issue was acceptable to not take any corrective actions. The team noted that in response to condition report CR-WF3-1998-1439, the licensee determined instrument uncertainty was not required to be considered for this application because the instrument was not significant to safety. In response to condition report CR-WF3-2003-2115, the licensee determined the intent of the surveillance was to demonstrate that leakage would be less than 200 standard cubic feet per minute and since the instrument uncertainty was less than 0.125 inches water gauge the intent of the surveillance requirement was satisfied. As a result the instrument uncertainty for the differential pressure instrumentation was not taken into account during periodic testing for Surveillance Requirement 4.7.6.5.a. (This issue is discussed in more detail in Report Section 4OA2 e.)

Example 3 - Noncited Violation 05000382/2003006-02: Inadequate Design Control of the Diesel Generator Starting Air System.

NRC inspection report 05000382/2003006 documented a design control violation with cross-cutting aspects related to inadequate problem evaluation. The licensee failed to effectively evaluate a deficiency related to the design of the emergency diesel generator. The licensee incorrectly determined that the five start capability of the emergency diesel starting air receivers was only a sizing criterion for the receivers and that maintaining receiver pressure greater than the original test pressure was not required for system operability. As a result of this inadequate evaluation, the receiver air pressure was not always maintained above the original minimum test pressure.

c. Effectiveness of Corrective Actions

(1) Inspection Scope The team reviewed approximately 135 condition reports to verify that corrective actions related to the issues were identified and implemented in a timely manner commensurate with safety, including corrective actions to address common cause or generic concerns.

The team reviewed corrective actions planned and implemented by the licensee and sampled specific technical issues to determine whether adequate decisions related to structure, system, and component operability were made.

In addition, the team reviewed a sample of those condition reports written to address NRC inspection findings to ensure that the corrective actions adequately address the issues as described in the inspection report writeups. The team also reviewed a sample of corrective actions closed to other condition reports and programs, such as work requests and engineering work requests, to ensure that the condition described was adequately addressed and corrected.

A listing of specific documents reviewed during the inspection is included in the attachment to this report.

(2) Assessment The team evaluated several occurrences where the licensee did not effectively address conditions adverse to quality and corrective actions taken were timely and appropriate.

These included two examples, identified by the team, where the licensee failed to take prompt corrective actions to resolve long-standing issues. The team also evaluated six other findings, identified by the NRC baseline inspection program at Waterford 3, since the last problem identification and resolution inspection that had cross cutting aspects related to prompt and effective corrective actions to resolve conditions adverse to quality.

Example 1 - Noncited Violation 05000382/2004006-01: Failure to Promptly Identify Inappropriate Assumption and Correct Control Room Operator Dose Analysis The team determined that the licensee failed to take prompt corrective actions to address a long term design deficiency. The licensee identified problems with the assumptions used to preform the control room operator dose analysis and initiated a condition report on March 9, 2004. The licensee took no immediate corrective actions to either revise the dose calculation or take required compensatory measures. The licensee determined that they would evaluate the results of the tracer gas testing scheduled for April 16, 2004 to re-evaluate the control room habitability assumptions and procedural guidance. (This issue is discussed in more detail in Report Section 4OA2 e.)

Example 2 - Noncited Violation 05000382/2004006-02: Failure to Promptly Correct Over-pressure Condition in Main Feed Water Isolation Valve Hydraulic Operating Systems The team determined that the licensee failed to take prompt corrective actions to address hydraulic operating system over-pressure conditions. Consequently, additional problems were experienced. First, on February 10, 2001, operators found that one FW-184A hydraulic accumulator thermal relief valve had failed to relieve pressure at its 5400 psig setpoint. The as-found system pressure was 5595 psig. The licensee did not promptly correct the problem, and on February 16, 2001, operators found pressure at 6080 psig, which was above the design pressure of 5900 psig. Second, on February 20, 2004, operators found FW-184B hydraulic operating pressure at 5983 psig, which was above the design limit of 5900 psig. The licensee did not properly address operability and did not promptly identify and correct the problem. As a result, the replacement valve also failed due to the same cause (gelled hydraulic oil), and the system was over-pressurized. (This issue is discussed in more detail in Report Section 4OA2 e.)

Example 3 - Noncited Violation 05000382/2003007-03: Ineffective Corrective Actions to Prevent Recurrence of Voiding Conditions.

NRC inspection report 05000382/2003007 documented ineffective corrective actions to prevent recurrence of voiding conditions. Operators failed to take corrective actions to fill and vent the low pressure safety injection as required by condition report CR-WF3-2002-0818, which resulted in the system being inoperable on another occasion due to voiding.

Example 4 - Noncited Violation 05000382/2003007-04: Ineffective Corrective Actions to Prevent Recurrence of Primary Water Stress Corrosion Cracking of Alloy 600 Material.

NRC inspection report 05000382/2003007 documented ineffective corrective actions to prevent cracking of alloy 600 material. The licensee did not establish replacement plans for susceptible Alloy 600 nozzles and had no inspection plans other than visual examinations to find leakage during the subsequent outage.

Example 5 - Noncited Violation 05000382/2003011-04: Inadequate Corrective Actions for Identified Emergency Lighting Inadequacies.

NRC inspection report 05000382/2003011 documented inadequate corrective actions for emergency lighting inadequacies. On May 15, 2002, a special test instruction was prepared to perform a full field test of emergency lighting as a corrective action for condition report CR-WF3-2000-0665. The test was postponed twice and on July 3, 2003 the licensee performed an evaluation to justify cancellation of the test. The maintenance rule function failure determination for condition report CR-WF3-2000-1206 did not result in a proactive corrective action to address multiple emergency lighting discharge test failures.

Example 6 - Noncited Violation 05000382/2004002-01: Inadequate Corrective Actions Affecting Feedwater Isolation Valves.

NRC inspection report 05000382/2004002 documented inadequate corrective actions to address age related failures. Failure to replace feedwater isolation valve actuator O-rings that were susceptible to age related failures, as a corrective action to condition report CR-WF3-2000-0628, resulted in the inoperability of a feedwater isolation valve seven months later.

Example 7 - Noncited Violation 05000382/2004002-02: Inadequate Maintenance Instructions Affecting Main Feedwater Isolation Valve.

NRC inspection report 05000382/2004002 documented a maintenance control violation with cross-cutting aspects related to inadequate corrective actions. Condition Report CR-WF3-2002-0624 identified a need for an operating and maintenance manual for the feedwater isolation valves. A vender manual was identified, but the condition report was closed before maintenance instructions were amended to include torque specifications for the feedwater isolation valve hydraulic operating system. As a result, an O-ring failed seven days after it was replaced.

Example 8 - Noncited Violation 05000382/2004002-05: Failure to Promptly Identify and Correct Emergency Diesel Generator Loading and Fuel Oil Consumption Analysis Deficiencies.

NRC inspection report 05000382/2004002 documented a design control violation with cross-cutting aspects related to inadequate corrective actions. Condition Report CR-WF3-2003-3088, which identified that the diesel generator fuel oil consumption analysis did not properly account for three discrepancies, was closed to condition report CR-WF3-2003-2758. CR-WF3-2003-2758 was closed without revising the fuel oil consumption analysis for the deficiencies noted in CR-WF3-2003-3088.

d. Assessment of Safety-Conscience Work Environment

(1) Inspection Scope The team interviewed more than 15 individuals from the licensees staff, representing a cross-section of functional organizations and supervisory and non-supervisory personnel. These interviews assessed whether conditions existed that would challenge the establishment of a safety-conscience work environment. The team interviewed the site employees concern program coordinator and noted that a separate coordinator was assigned to the contractor security force. The team also reviewed the most recent employee survey performed by the licensee to evaluate the health of the plants safety culture and the employees concerns program.

(2) Assessment The team concluded that a positive safety-conscience work environment exists at Waterford 3. The team determined that employees and contractors feel free to raise safety concerns to their supervision or bring concerns to the employees concern program. The team determined that licensee management is receptive to employee concerns and is willing to address issues raised by the latest safety culture survey.

e. Specific Issues Identified During This Inspection

(1) Inspection Scope During this assessment the team performed the inspections scoped in Sections 4OA2 a.(1), 4OA2 b.(1), 4OA2 c.(1),and 4OA2 d.(1) above.

(2) Finding Details Noncited Violation 05000382/2004006-01: Failure to Promptly Identify Inappropriate Assumption and Correct Control Room Operator Dose Analysis

Introduction.

The team identified a noncited violation of 10 CFR 50, Appendix B, Criterion XVI, for the failure to promptly identify and correct a condition adverse to quality. Specifically, on multiple occasions the licensee failed to identify and correct an inappropriate value of the unfiltered inleakage parameter used to calculate the control room operator dose for design basis accident conditions involving radiological releases.

This failure resulted in significantly underestimating the actual dose to operators.

Description.

In April 2004, the licensee performed tracer gas testing of the control room envelope in response to concerns addressed in NRC Generic Letter 2003-001. Results of the tracer gas test demonstrated that assumed unfiltered inleakage values utilized by the control room operator dose calculation were exceeded for design basis accident conditions. The team reviewed the design of the control room ventilation system and noted that following a high radiation condition or safety injection actuation event the control room would automatically isolate, but would not automatically pressurize. The team noted that control room operator dose calculation, EC-S96-011, assumed 13 cfm of unfiltered inleakage. This value was assumed based on the control room being in a pressurized mode of operation. The team questioned the licensee why they believed 13 cfm would adequately model the event since the control room did not automatically pressurize as assumed by the analysis. The licensee stated the analysis was considered a bounding analysis since they had no data to indicate otherwise. The team was concerned that the analysis was not bounding since unfiltered inleakage would be higher when the control room was isolated, as compared to isolated and pressurized.

The team reviewed the licensees original Final Safety Analysis Report and noted that the control room ventilation system was designed to automatically isolate and pressurize following a safety injection actuation signal or a control room high radiation condition. The control room dose calculation used the appropriate unfiltered inleakage assumption for this design. In 1983, the licensee changed the design of the control room ventilation system to automatically isolate following a safety injection actuation signal or a control room high radiation condition without automatic pressurization. This design change was performed without changing the unfiltered inleakage assumption for the control room dose calculation to account for a higher assumed unfiltered inleakage from adjacent areas. NUREG-0800, Standard Review Plan, Section 6.4, specifies that unfiltered inleakage for control rooms that isolate without pressurization use an unfiltered inleakage value of one half the amount of air required to pressurize the envelope to one eighth inch water gauge. This specification required the licensee to assume approximately 100 cfm of unfiltered inleakage into the control room for the isolate mode of operation. The team noted that the control room dose calculation was revised in 1994 and 1998. Both of these revisions also failed to identify that the unfiltered inleakage criterion was inappropriate. The team determined that these were missed opportunities for the licensee to identify this deficient condition.

The team performed a review of the licensees corrective action database and identified that a system engineer had initiated a condition report on March 9, 2004, related to this same deficiency. Condition Report CR-WF3-2004-00725, stated, in part, that a potential adverse condition was discovered, Waterford 3's current design basis for control room habitability during a radiological emergency assumes the control room to be in the pressurized mode contrary to design basis documents. The licensee took no immediate corrective actions to either revise the dose calculation or take required compensatory measures. The licensee determined that they would evaluate the results of the tracer gas testing to re-evaluate the control room habitability assumptions and procedural guidance. The team determined that this was a missed opportunity to correct this deficient condition.

Analysis.

The deficiency associated with this finding was the failure to promptly identify and correct a condition adverse to quality. Specifically, on multiple occasions the licensee failed to identify and correct an inappropriate unfiltered inleakage parameter used to calculate the control room operator dose for design basis accident conditions involving radiological releases. This failure resulted in underestimating the dose to operators following design basis accident conditions involving radiological releases.

This finding was greater than minor because it affected the barrier integrity cornerstone objective related to design control of the control room envelope. This finding was evaluated using Inspection Manual Chapter 0609, Significance Determination Process, Phase 1 worksheet under the containment barriers cornerstone. The finding was determined to be of very low safety significance because the deficiency only affected the radiological barrier function provided for the control room.

Enforcement.

10 CFR 50, Appendix B, Criterion XVI, Corrective Action, states, in part, that measures be established to assure that conditions adverse to quality are promptly identified and corrected. The failure to promptly identify and correct the use of an inappropriate unfiltered inleakage parameter affecting the control room operator dose analysis is a violation of 10 CFR 50, Appendix B, Criterion XVI. Because the failure to promptly identify and correct the analysis was of very low safety significance and has been entered into the licensees corrective action program as Condition Report CR-WF3-2004-1403, this violation is being treated as a noncited violation, consistent with Section VI.A of the NRC Enforcement Policy:

(NCV 05000382/2004006-01) Failure to Promptly Identify Inappropriate Assumption and Correct Control Room Operator Dose

Analysis.

Noncited Violation 05000382/2004006-02: Failure to Promptly Correct Over- pressure Condition in Main Feed Water Isolation Valve Hydraulic Operating Systems

Introduction:

The team identified a 10 CFR 50, Appendix B, Criterion XVI, noncited violation for situations where the licensee failed to promptly correct conditions adverse to quality associated with the main feed isolation valve hydraulic actuating systems. In two cases, the licensee failed to promptly correct high pressure conditions where the hydraulic actuator thermal relief valves failed to properly function. Consequently, the hydraulic portion of the valve actuator experienced repetitive over-pressure conditions.

In one case, engineers failed to properly address system operability and, for a two-week period, actual valve operability was unknown. Contributors to the problems included 1) engineers did not properly understand all aspects of system design, and 2) the system did not have readily available pressure indication or high pressure alarms.

Discussion: The main feed isolation Valves FW-184A and FW-184B automatically close for containment isolation and to limit reactor coolant system cooldown during certain design basis accidents. Acceptable valve closure times range from 2.3 to 5.0 seconds.

The lower closure limit is to preclude a water-hammer, which can be induced by fast valve closure. Each main feed isolation valve hydraulic operating system is equipped with two hydraulic accumulators. Both accumulators are required for proper valve operation. In order to limit valve speed, accumulator pressure is limited to 5900 psig.

Each accumulator is equipped with a thermal relief valve to prevent over-pressurization.

The team identified two examples where the licensee failed to promptly correct conditions adverse to quality and, as a result, additional problems were experienced.

• On February 10, 2001, operators found that one FW-184A hydraulic accumulator thermal relief valve had failed to relieve pressure at its 5400 psig setpoint. The as-found system pressure was 5595 psig. The malfunctioning thermal relief valve was a condition adverse to quality. The licensee did not promptly correct the problem and on February 16, 2001, operators found pressure at 6080 psig, which was above the design pressure of 5900 psig. The licensee ultimately identified that a retaining clip was not properly installed in the relief valve.

• On February 20, 2004, operators found FW-184B hydraulic operating pressure at 5983 psig, which was above the design limit of 5900 psig. The team identified that the licensee did not perform an adequate operability evaluation and did not promptly correct the condition adverse to quality. In the operability assessment, the licensee stated that accumulator pressure was reduced therefore FW-184B was operable. However, the operability assessment did not address the potential for the system to over-pressurize again. The licensee did not start additional monitoring of hydraulic pressure until March 4, 2004, and then only on an infrequent basis. The status of the hydraulic system between February 20 and March 4 is unknown. On March 25, 2004, the licensee replaced the thermal relief valve. In addition, the licensee did not promptly identify the cause of the February 20, 2004, thermal relief valve malfunction and on March 29, 2004, the replacement relief valve also failed. The failures were caused by debris (gelled hydraulic fluid) in the system. The licensee did not have the first relief valve tested until April 15, 2004, almost two months after its failure. After the second relief valve failure, the licensee performed more frequent pressure monitoring until the valve was replaced on May 18, 2004.

The team determined that poor understanding of system performance contributed to the continuing problems. Engineers did not fully understand what was causing the over-pressure conditions. Engineers believed that hydraulic system temperature alone was the cause for the system pressure increases and set up monitoring based on outside ambient temperature changes. The team determined that temperature alone could not explain the system pressure behavior. According to the hydraulic system vendor manual, pressure increases approximately 10 psig per degree Fahrenheit rise.

The team observed that system pressure was not predictable based on this simple relationship. Therefore, there were other factors affecting hydraulic pressure that the licensee did not understand. In some cases, hydraulic system pressure increased as ambient temperatures lowered.

The team also noted that records were not available to identify all of the instances where hydraulic pressure may have exceeded its design limit. The hydraulic system accumulators do not have permanently installed pressure gages and no alarm warns operators of a potential over-pressure condition. Instead, craftsmen installed temporary gages each time that pressure monitoring was desired. Consequently, the licensee infrequently checked pressure, and typically only when a problem was suspected. In some instances, over-pressure conditions were identified after a low pressure alarm was received on one of the hydraulic system accumulators.

At the time of the inspection, the licensee was planning to add a high pressure alarm to the system by July 2004. As additional corrective measures, the licensee started monitoring accumulator pressures more frequently.

Analysis:

The team determined that the issue was more than minor in significance because it affected the mitigating systems cornerstone objective to ensure the availability of systems that respond to initiating events. However, the finding was determined to be of very low risk significance because the issue:

(1) was not a design or qualification deficiency;

(2) did not result in the loss of a safety system;

(3) did not represent an actual loss of a safety function of a single train for greater than its technical specification allowed outage time;

(4) did not represent an actual loss of safety function of one or more non-Technical Specification trains of equipment designated as risk significant per 10 CFR 50.65 for greater than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />; and

(5) was not potentially risk significant due to a seismic, fire, flooding, or severe weather initiating event.

Enforcement:

The team identified a noncited violation of 10 CFR 50, Appendix B, Criterion XVI, which requires the licensee to take prompt measures to correct conditions adverse to quality. The valve over-pressurization events, described above, are considered conditions adverse to quality. In each case, the licensee failed to promptly correct the conditions and additional consequences were experienced. Because the failure to promptly identify and correct the over-pressure condition was of very low safety significance and has been entered into the licensees corrective action program as condition reports CR-WF3-2004-1533, 1540 and 1551, this violation is being treated as a noncited violation, consistent with Section VI.A of the NRC Enforcement Policy:

(NCV 05000382/2004006-02) Failure to Promptly Correct Overpressure Condition in Main Feed Water Isolation Valve Hydraulic Operating Systems.

Noncited Violation 05000382/2004006-03: Failure to Promptly Correct a Known Deficient Condition Involving the Failure to Account for Instrument Uncertainty to Satisfy Technical Specification Surveillance Requirement 4.7.6.5.a.

Introduction.

The team identified a Green noncited violation of 10 CFR 50, Appendix B, Criterion XVI, for the failure to promptly identify and correct a condition adverse to quality. Specifically, on multiple occasions the licensee failed to correct a known deficient condition involving the failure to account for instrument uncertainty to satisfy Technical Specification Surveillance Requirement 4.7.6.5.a. This failure potentially affects the ability of the control room envelope to perform its design function with respect to protecting operators from postulated design basis accidents resulting in radiological releases.

Discussion. The team reviewed Condition Reports CR-WF3-1998-1439 and CR-WF3-2003-2115. Condition Report CR-WF3-1998-1439 discusses that during a review of design basis review open items it was noted that the differential pressure instruments used to satisfy Technical Specification 4.7.6.5.a have insufficient allowance between the 0.125 inch water gauge requirement and the procedure acceptance criteria (between 0.125 and 0.130 inch water gauge) to account for instrument uncertainty of approximately 0.2 inch water gauge. The team noted that the licensee determined instrument uncertainty was not required to be considered for this application because the instrument was not significant to safety. CR-WF3-2003-2115, written by electrical engineering design staff, discusses that due to the large uncertainty associated with the differential pressure instrument loop, measuring and test equipment should be used to satisfy the Surveillance Requirement 4.7.6.5.a requirements. A rigorous engineering calculation was performed demonstrating the instrument uncertainty to be 0.113 inch water gauge as read on the control room indicator, and 0.105 inch water gauge as read on the plant monitoring computer. The surveillance procedure allows use of either reading. The team noted that the licensee determined the intent of the surveillance was to demonstrate that leakage would be less than 200 standard cubic feet per minute with the control room envelope at a positive pressure above 0.0 inch water gauge, and since the instrument uncertainty was less than 0.125 inches water gauge the intent of the surveillance requirement was satisfied.

The team reviewed the licensees Technical Specifications Surveillance Requirement 4.7.6.5.a and the Final Safety Analysis Report. Surveillance Requirement 4.7.6.5.a states, The control room envelope isolation and pressurization boundaries shall be demonstrated operable at least once per 18 months by verifying that the control room envelope can be maintained at a positive pressure of greater than or equal to 1/8 inch water gauge relative to the outside atmosphere with a make-up air flowrate less than or equal to 200 cfm during system operation. Final Safety Analysis Report Chapter 6.4.3.2, Habitability Systems, Leak Tightness, states, in part, An acceptance test will be performed to verify the adequacy of the air makeup rate to maintain positive pressure inside the control room envelope of at least +0.125 inch water gauge. This section also states, Since the leakage analysis shows a gross leakage rate of less than 0.06 volume changes per hour, gross leakage will be verified by periodic testing as described in Regulatory Guide 1.95. NRC Regulatory Guide 1.95, Protection of Nuclear Power Plant Control Room Operators Against an Accidental Chlorine Release, states in part, The gross leakage characteristics of the control room should be determined by pressurizing the control room to 1/8-inch water gage and determining the pressurization flow rate (The use of a higher pressure differential is acceptable provided the flow rate is conservatively adjusted to correspond to 1/8-inch water gage).

Based on this information the team determined that the intent of the surveillance requirement was to validate the leak-tightness of the control room envelope at a minimum of 0.125 inch water gauge. The team noted that the Technical Specifications, Final Safety Analysis Report and NRC Regulatory Guide 1.95 consistently state the intent of the test is to validate that at least 0.125 inch water gauge is obtained during the control room pressurization test. The team determined that the licensees position that the intent of the surveillance is to demonstrate that leakage will be less than 200 standard cubic feet per minute with the control room envelope at a positive pressure above 0.0 inches water gauge was not appropriate since it reduces the margin of safety as described in the licensees Technical Specifications and Final Safety Analysis Report.

The team determined that the failure to correct a known deficient condition to account for instrument uncertainty for satisfying Technical Specification Surveillance Requirement 4.7.6.5.a was a violation of 10 CFR, Part 50, Appendix B, Criterion XVI, Corrective Action.

Analysis.

The deficiency associated with this finding was the failure to correct a condition adverse to quality. Specifically, on multiple occasions the licensee failed to correct a known deficient condition involving the failure to account for instrument uncertainty to satisfy Technical Specification Surveillance Requirement 4.7.6.5.a. This failure potentially affects the ability of the control room envelope to perform its design function with respect to protecting operators from postulated design basis accidents resulting in radiological releases. This finding was greater than minor because it affected the barrier integrity cornerstone objective related to maintaining the barrier function of the control room envelope. This finding was evaluated using Inspection Manual Chapter 0609, Significance Determination Process, Phase 1 worksheet under the containment barriers cornerstone. The finding was determined to be of very low safety significance because the deficiency only affected the radiological barrier function provided for the control room.

Enforcement.

10 CFR 50, Appendix B, Criterion XVI, Corrective Action, states, in part, that measures be established to assure that conditions adverse to quality are promptly identified and corrected. The failure to promptly identify and correct a known deficient condition involving the failure to account for instrument uncertainty to satisfy Technical Specification Surveillance Requirement 4.7.6.5.a. is a violation of 10 CFR 50, Appendix B, Criterion XVI. Because the failure to promptly identify and correct the analysis was of very low safety significance and has been entered into the licensees corrective action program as Condition Report CR-WF3-2004-1561, this violation is being treated as a noncited violation, consistent with Section VI.A of the NRC Enforcement Policy: (NCV 05000382/2004006-03), Failure to Promptly Identify and Correct a Known Deficient Condition Involving the Failure to Account for Instrument Uncertainty to Satisfy Technical Specification Surveillance Requirement 4.7.6.5.a.

4OA6 Exit Meeting

The team discussed the findings with you and other members of the licensees staff on May 21, 2004. Licensee management did not identify any materials examined during the inspection as proprietary.

ATTACHMENT 1 PARTIAL LIST OF PERSONS CONTACTED Licensee L. Borel, Senior Quality Assurance Engineer E. Brauner, Acting Manager, System Engineering R. Brian, Director, Engineering J. Burke, Supervisor, Quality Assurance W. Campbell, Chief Operating Officer, Entergy Nuclear South C. DeDeaux, Licensing Engineer R. Dodds, Vice Presidents Staff C. Fugate, Manager, Operations T. Gaudet, Equipment Reliability Lead Engineer A. Harris, Manager, Engineering Projects J. Holman, Manager, Nuclear Engineering B. Huston, Acting Manager, Nuclear Safety and Assurance J. Laque, Manager, Maintenance B. Lindsey, Operations Control Room Supervisor D. Litoff, Operations Shift Technical Advisor R. Murillo, Senior Staff Engineer R. Osborne, Manager, Engineering Programs and Components R. Peters, Manager, Planning, Scheduling and Outages O. Pipkins, Licensing Engineer R. Porter, Superintendent, Mechanical Maintenance J. Rachal, Design Engineer J. Ridgel, Manager, Corrective Action Program T. Schreckengast, Operations Shift Manager K. Walsh, General Manager, Plant Operations R. Williams, Licensing Engineer J. Venable, Vice President, Operations NRC W. Jones, Chief, Project Branch E, Division of Reactor Projects, Region IV ITEMS OPENED AND CLOSED Opened and Closed 05000382/2004006-01 NCV Failure to Promptly Identify Inappropriate Assumption and Correct Control Room Operator Dose Analysis (Section 4OA2 e.)05000382/2004006-02 NCV Failure to Promptly Correct Over-pressure Condition in Main Feed Water Isolation Valve Hydraulic Operating Systems (Section 4OA2 e.)05000382/2004006-03 NCV Failure to Promptly Correct a Known Deficient Condition Involving the Failure to Account for Instrument Uncertainty to Satisfy Technical Specification Surveillance Requirement 4.7.6.5.a. (Section 4OA2 e.)

DOCUMENTS REVIEWED

PLANT PROCEDURES

Procedure Title Revision

CE-002-030 Maintaining Diesel Fuel Oil 8

DC-115 ER Response Development 4

LI-102 Corrective Action Process 4

LI-104 Self Assessment and Benchmark Process 5

LI-118 Root Cause Analysis Process 0

MI-005-644 Feedwater Isolation Valve A or B Operations Check 7

OE-100 Operating Experience Program 1

OP-100-0014 Technical Specification and Technical Requirements 13

Compliance

OP-902-002 Loss of Coolant Accident Recovery 9

OP-902-003 Loss of Offsite Power / Loss of Forced Circulation Recovery 4

OP-902-009 Standard Appendices 1

OP-903-046 Emergency Feedwater Pump Operability Check 15

OP-903-110 RAB Fluid Systems Leak Test 5

UNT-007-011 Section 6.7, Operability and Reportability Assessment 7

Guidance

UNT-005-004 Temporary Alteration Control 16

ENGINEERING REQUESTS

Number Title Revision/Date

ER-W3-98- Affects of uncontrolled maximum EFW on accident June 21, 2002

0764-0001 situations

ER-W3-00- RC-Reactor Coolant , CVC-Chemical and Volume Control Revision 0

0106-002 System

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ER-W3-00- Insulation Removal for HPSI November 2,

0896-00-00 2000

ER-W3-01- Safeguards Room B Safety Related Equipment Evaluation May 1. 2001

299-00-00

CALCULATIONS

Number Title Revision/Date

EC-P98-001 MFIV Closure Time FW Transient Analysis and February 2, 1999

Evaluation of FW Piping due to Check Valve Slam

EC-M98-003 Design Basis for Feedwater Isolation Valves FW-184A March 8, 2004

& FW-184B

EC-M101-001 Seismic and Weak Link Analysis for Feedwater Isolation February 2, 2001

Valves

EC-M02-001 Minimum Required EFW Pumps Discharge Pressure January 31, 2002

during Recirculation

EP-S-001-W Steam Generator ECT Data Analysis for Waterford 3 Revision 01

MNQ-10-1 Emergency Feedwater System Head Curves February 2, 2004

MPR-2299 Replacement Interval and Shelf Life of Electrolytic Revision 0

Capacitors Used in Woodward Controls

CONDITION REPORTS, CR-WF3-

1998-0792 2001-0352 2001-1232 2002-0953 2003-0090

1998-0844 2001-0410 2001-1341 2002-0987 2003-0140

1998-1439 2001-0433 2001-1360 2002-1002 2003-0147

1999-1208 2001-0436 2001-1364 2002-1230 2003-0199

2000-0148 2001-0456 2002-0178 2002-1422 2003-0241

2000-0432 2001-0500 2002-0220 2002-1426 2003-0384

2000-0341 2001-0512 2002-0224 2002-1516 2003-0395

2000-0470 2001-0649 2002-0297 2002-1587 2003-0977

2000-1270 2001-0748 2002-0339 2002-1898 2003-1069

2000-1564 2001-0831 2002-0390 2002-1949 2003-1242

2001-0109 2001-0829 2002-0408 2002-1950 2003-1303

2001-0191 2001-0852 2002-0448 2002-2147 2003-1317

2001-0225 2001-0853 2002-0804 2003-0041 2003-1345

2001-0247 2001-0988 2002-0824 2003-0062 2003-1441

2001-0317 2001-1110 2002-0907 2003-0089 2003-1528

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2003-1679 2003-2555 2003-3716 2004-0156 2004-0861

2003-1687 2003-2666 2003-3721 2004-0304 2004-0862

2003-1794 2003-2832 2003-3740 2004-0311 2004-0874

2003-1942 2003-2972 2003-3803 2004-0477 2004-0973

2003-2044 2003-3110 2003-3837 2004-0524 2004-0987

2003-2064 2003-3130 2003-3891 2004-0551 2004-1433

2003-2075 2003-3164 2003-3997 2004-0584 2004-1518

2003-2097 2003-3260 2004-0020 2004-0608 2004-1521

2003-2115 2003-3263 2004-0026 2004-0725 2004-1533

2003-2147 2003-3280 2004-0050 2004-0784 2003-0033

2003-2495 2003-3371 2004-0053 2004-0800 2003-0030

2003-2502 2003-3592 2004-0126 2004-0827 2003-1614

2003-2554

ROOT CAUSE ANALYSIS REPORTS FOR CR-WF3-

2001-0317 2002-0339 2003-0062 2003-3891 2004-759 2004-1011

LEARNING ORGANIZATION CONDITION REPORTS

LO-OPX-2002-0323 LO-OPX-2003-0217 LO-OPX-2004-0033 LO-WLO-2004-0011

LO-OPX-2003-0140 LO-OPX-2003-0203 LO-OPX-2004-0097

WORK ORDERS

01149460 00406063 00413294 00418106 00427487

00016935 00406123 00413294 00418833 50010593

00028163 00408858 00414583 00421755 50010596

00402623 00411419 00416489 00423685 50010598

00405942 00411422 00417942 00426301 50093392

MAINTENANCE ACTION ITEMS

24076 23329 24077 32748 34888 34904

OTHER

Diesel Fuel Oil Storage Tank Administrative and Alarm Limits February 14, 2004

2004 Top Ten Equipment Reliability Issues April 27, 2004

-6-

CFR 50.59 Screening for Compensatory Measures for January 9, 2003

Switchgear Ventilation Damper SVS-102

Safety Injection System Description Revision 7

Emergency Feedwater System Description Revision 8

Emergency Feedwater System Health Report May 5, 2004

Emergency Feedwater Design Basis Document Revision 2-7

Entergy Nuclear South Alloy 600 Project, Long Range Plan February 9, 2004

Waterford 3 Safety Culture Survey Summary Report May 12, 2004

ENS Corrective Action Program Quality Assurance Audit Report June 10, 2003

Problem Identification & Resolution Assessment March 12, 2004

IST Trend Data for 9 pumps and 11 valves

Waterford 3 Daily Plant Status Reports

1st Quarter 2004 Waterford 3 Quarterly Trend Report

Entergy Nuclear South Condition Report Initiation Guidance

Pamphlet

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INITIAL MATERIAL REQUEST

INITIAL INFORMATION REQUEST FROM WATERFORD 3

FOR PI&R INSPECTION (Report Number 05000382/2004006)

The inspection will cover the period of October 2002 to March 2004. The information may be

provided in either electronic or paper media or a combination thereof. Information provided in

electronic media may be in the form of CDs, or 31/2 inch floppy disks. The agencys text editing

software is Corel WordPerfect 8, Presentations, and Quattro Pro; however, we have document

viewing capability for MS Word, Excel, Power Point, and Adobe Acrobat (.pdf) text files.

Please provide the following information to Peter Alter by March 29, 2004 at the Resident

Inspector Office at Waterford-3

All procedures governing or applying to the corrective action program, including the

processing of information regarding generic communications and industry operating

experiences

Procedures and descriptions of any informal systems, used by engineering, operations,

maintenance, security, training, and emergency planning for issues below the threshold

of the formal corrective action program

A searchable table of all corrective action documents [condition reports (CRs)] that were

initiated or closed during the period, include CR number, description of issue and

significance classification

Either annotate on the above list or a separate list of all CRs associated with:

(1) Human performance issues

(2) Emergency preparedness issues

(3) Response to 10 CFR 21 reports

A separate list of all CRs closed to other programs, such as MAIs/WOs, ERs, etc.

A copy of each Significant Event Review Team Report and Root Cause Analysis Report

for the period (not necessarily the whole CR)

Copies of CRs (for the period) associated with non-escalated (no response required) or

noncited violations for the period

Copies of CRs (for the period) associated with repetitive problems or issues

Copies of CRs (for the period) associated with ineffective or untimely corrective actions

List of all self assessments or QA assessments/audits for the period

All corrective action program reports or metrics used for tracking effectiveness of the

corrective action program for the period

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All quality assurance audits and surveillances, and functional self assessments of

corrective action activities completed for the period

Control room logs for the year 2003

Security event logs for the year 2003

Radiation protection event logs for the year 2003

List of risk significant systems from W3 PRA/PSA, based on risk achievement worth

(RAW) and "0% availability CDF"

Searchable list of all maintenance action items/work orders for the period

List of all SSCs placed in or removed from the maintenance rule a(1) category for the

period

All corrective action documents related to the following industry operating experience

generic communications:

NRC Bulletins

NRC BULLETIN 2002-001 Reactor Pressure Vessel Head Degradation and Reactor

Coolant Pressure Boundary Integrity

NRC Information Notices:

NRC INFORMATION NOTICE 2004-001 Auxiliary Feedwater Pump Recirculation Line

Orifice Fouling - Potential Common Cause Failure

NRC INFORMATION NOTICE 2003-019 Unanalyzed Condition of Reactor Coolant

Pump Seal Leakoff Line During Postulated Fire Scenarios or Station Blackout

NRC INFORMATION NOTICE 2003-013 Steam Generator Tube Degradation at Diablo

Canyon

NRC INFORMATION NOTICE 2003-011 Leakage Found on Bottom-Mounted

Instrumentation Nozzles

NRC INFORMATION NOTICE 2003-008 Potential Flooding Through Unsealed Concrete

Floor Cracks

NRC INFORMATION NOTICE 2003-005 Failure to Detect Freespan Cracks in PWR

Steam Generator Tubes

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NRC INFORMATION NOTICE 2003-002 Recent Experience With Reactor Coolant

System Leakage And Boric Acid Corrosion

NRC INFORMATION NOTICE 2002-034 Failure of Safety-Related Circuit Breaker

External Auxiliary Switches at Columbia Generating Station

Attachment