IR 05000382/2001009

From kanterella
Jump to navigation Jump to search
IR 05000382/2001-009; on 12/30/01-03/30/02; Entergy Operations, Inc.; Waterford Steam Electric Station, Unit 3; Integrated Resident & Regional Report; Maintenance Risk Assessments and Emergent Work Evaluation; Permanent Plant Modifications
ML021160080
Person / Time
Site: Waterford Entergy icon.png
Issue date: 04/25/2002
From: William Jones
NRC/RGN-IV/DRP/RPB-E
To: Venable J
Entergy Operations
References
IR-01-009
Download: ML021160080 (33)
v • d • e


Text

ril 25, 2002

SUBJECT:

NRC INSPECTION REPORT 50-382/01-09

Dear Mr. Venable:

On March 30, 2002, the NRC completed an inspection at your Waterford Steam Electric Station, Unit 3. The enclosed report documents the inspection findings which were discussed on January 18, March 1, and April 3, 2002 , with you and other members of your staff.

This 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.

Within these areas, the inspection consisted of selected examination of procedures and representative records, observations of activities, and interviews with personnel.

Based on the results of this inspection, the NRC has identified three issues that were evaluated under the risk significance determination process as having very low safety significance (Green). The NRC has also determined that violations are associated with these issues.

These violations are being treated as noncited violations (NCVs), consistent with Section VI.A of the Enforcement Policy. These NCVs are described in the subject inspection report. If you contest the violation or significance of these NCVs, you should provide a response within 30 days of the date of this inspection report, with the basis for your denial, to the U.S. Nuclear Regulatory Commission, ATTN: Document Control Desk, Washington, DC 20555-0001, with copies to the Regional Administrator, 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.

Since September 11, 2001, Waterford 3 has assumed a heightened level of security based on a series of threat advisories issued by the NRC. Although the NRC is not aware of any specific threat against nuclear facilities, the heightened level of security was recommended for all nuclear power plants and is being maintained due to the uncertainty about the possibility of additional terrorist attacks. The steps recommended by the NRC include increased patrols, augmented security forces and capabilities, additional security posts, heightened coordination with local law enforcement and military authorities, and limited access of personnel and vehicles to the site. In February 2002, the NRC issued an order to all commercial power plants to implement interim compensatory measures for the generalized high-level threat environment.

Entergy Operations, Inc. -2-Some of the requirements formalize a series of security measures that NRC licensees had taken in response to advisories issued by the NRC. The order also imposes additional security requirements which have emerged from the ongoing security review.

In accordance with 10 CFR 2.790 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 (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).

Should you have any questions concerning this inspection, we will be pleased to discuss them with you.

Sincerely,

/RA/

William B. Jones, Chief Project Branch E Division of Reactor Projects Docket: 50-382 License: NPF-38

Enclosure:

NRC Inspection Report 50-382/01-09

REGION IV==

Docket: 50-382 License: NPF-38 Report: 50-382/01-09 Licensee: Entergy Operations, Inc.

Facility: Waterford Steam Electric Station, Unit 3 Location: Hwy. 18 Killona, Louisiana Dates: December 30, 2001, through March 30, 2002 Inspectors: T. R. Farnholtz, Senior Resident Inspector G. F. Larkin, Resident Inspector M. F. Runyan, Senior Reactor Inspector, DRS C. J. Paulk, Senior Reactor Inspector, DRS P. A. Goldberg, Reactor Inspector, DRS J. B. Nicholas, Ph.D., Senior Health Physicist, DRS Accompanying Personnel: J. Taylor, Reactor Inspector, DRS Approved By: W. B. Jones, Chief, Project Branch E

SUMMARY OF FINDINGS Waterford Steam Electric Station, Unit 3 NRC Inspection Report 50-382/01-09 IR05000382-01-09; on 12/30/01-03/30/02; Entergy Operations, Inc.; Waterford Steam Electric Station, Unit 3; Integrated Resident & Regional Report; Maintenance Risk Assessments and Emergent Work Evaluation; Permanent Plant Modifications.

The inspection was conducted by resident inspectors, two senior reactor inspectors, a reactor inspector, and a senior health physicist. The inspection identified three Green issues. The significance of most findings is indicated by their color (Green, White, Yellow, Red) using IMC 0609, Significance Determination Process (SDP). The NRCs program for overseeing the safe operation of commercial nuclear power reactors is described at its Reactor Oversight Process website at http://www.nrc.gov/NRR/OVERSIGHT/index.html. Findings for which the SDP does not apply are indicated by No Color or by the severity level of the applicable violation.

Cornerstone: Mitigating Systems

  • Green. The inspectors identified a violation of Technical Specification 6.8.1 for the failure to meet the reactivity management program requirements during the performance of maintenance on Charging Pump A. The work package for Charging Pump A did not include a completed reactivity management checklist used to document the reactivity management program screening. The reactivity management program requires that work on specified systems such as the charging system be screened for the potential of an inadvertent reactivity change. Subsequent to this finding, the licensee performed a self-assessment to determine the extent of this condition.

Additional issues with the reactivity management program were identified. The inspectors considered these issues to be programmatic in nature in that the program requirements were not being met in all cases for maintenance activities. This violation is being treated as a noncited violation, consistent with Section VI.A.1 of the NRC Enforcement Policy and is in the licensees corrective action program as Condition Reports 2002-0169 and -0476.

This violation was more than minor because it could be reasonably viewed as a precursor to a more significant event due to the potential for an unplanned reactivity excursion and could affect the function of the charging pump or other reactivity management systems. This issue was determined to be of very low safety significance because there was no inadvertent reactivity change (Section 1R13.5).

  • Green. The inspectors identified a violation of Technical Specification 6.8.1 for the failure to perform corrective maintenance on a reactor trip circuit breaker in accordance with established procedures. During installation of a reactor trip circuit breaker, the breaker unexpectedly closed as it was being placed into service. The licensee performed troubleshooting and repair activities on the breaker and subsequently placed the breaker in service. No record of the troubleshooting or repair activities was made, resulting in an inability to independently verify the specifics of the problem or provide for traceability of parts used, as required by corrective maintenance procedures. This is

-2-being treated as a noncited violation consistent with Section VI.A.1 of the NRC Enforcement Policy and is in the licensees corrective action program as Condition Report 2002-0382.

The safety significance of this violation was determined to be more than minor because there was a credible impact on safety by not performing corrective maintenance in accordance with established procedures on safety-related equipment (reactor trip circuit breaker), which could affect the operability, availability, reliability, or function of the reactor protection system. Using the reactor safety significance determination process, the violation was determined to have very low safety significance because the reactor trip breakers would have functioned if required (Section 1R13.6).

Cornerstone: Barrier Integrity

  • Green. The inspectors identified a violation of Criterion III of Appendix B to 10 CFR Part 50 for a design change that failed to fully consider the requirements of Article NC-7000, Protection Against Overpressure, of Section III in the ASME Boiler and Pressure Vessel Code, 1971 Edition through Winter 1972 Addenda. This failure resulted in the approval to install a relief valve with a setpoint greater than the design pressure in a section of pipe in a containment penetration that is normally isolated with entrained fluid. This design change had a credible impact on safety because the design change directed the installation of a relief valve with a set pressure greater than the design pressure allowed by the ASME Code. This design change also could affect the integrity of the containment barrier as a result of not providing overpressure protection such that the design pressure of any component within the boundary would not be exceeded. This violation is being treated as a noncited violation consistent with Section V1.A.1 of the NRC Enforcement Policy and is in the licensees corrective action program as Condition Report CR-WF3-2002-0079.

This issue was determined to be of very low safety significance because the modification was not installed in the plant and this design did not represent: a degradation of the radiological barrier function provided for the control room, or auxiliary building, or spent fuel pool; a degradation of the barrier function of the control room against smoke or a toxic atmosphere; or an actual open pathway in the physical integrity of reactor containment or an actual reduction of the atmospheric pressure control function of the reactor containment (Section 1R17b.).

Report Details Summary of Plant Status: The plant was at 100 percent power at the beginning of this inspection period. On March 6, 2002, operators commenced a plant power coastdown in preparation for Refueling Outage 11. On March 23, the plant was shut down from approximately 87 percent power to commence the refueling outage and remained in that condition for the remainder of the inspection period.

1 REACTOR SAFETY Initiating Events, Mitigating Systems, Barrier Integrity, Emergency Preparedness 1R02 Evaluation of Changes, Tests, or Experiments (71111.02)

a. Inspection Scope The inspectors reviewed a selected sample of nine safety evaluations, listed in the attachment to this report, to verify that the licensee had appropriately considered the conditions under which the licensee may make changes to the facility or procedures or conduct tests or experiments without prior NRC approval.

The inspectors reviewed a selected sample of 11 safety evaluation screenings, listed in the attachment to this report, in which the licensee determined that safety evaluations were not required to ensure that the licensees exclusion of a full evaluation was consistent with the requirements of 10 CFR 50.59, Evaluations of Changes, Tests, or Experiments.

The inspectors reviewed a selected sample of four condition reports, listed in the attachment to this report, initiated by the licensee to address problems or deficiencies associated with its implementation of 10 CFR 50.59 requirements to verify that appropriate corrective actions had been accomplished.

b. Findings No findings of significance were identified.

1R04 Equipment Alignment (71111.04)

a. Inspection Scope The inspectors reviewed the following system alignments during this inspection period:

  • Component Cooling Water Train A: On February 4, 2002, the inspectors completed a review and partial system walkdown of Component Cooling Water Train A, which was aligned in standby while Component Cooling Water Train B was out of service for scheduled replacement of air accumulator tanks associated with Valves CC-134B and -135B. The review included the Updated Final Safety Analysis Report and Operations Procedure OP-002-003, Component Cooling Water System, Revision 13.

-2-

  • Auxiliary Component Cooling Water Train A: On February 4, 2002, the inspectors walked down and observed the mechanical and electrical alignment of critical portions of Auxiliary Component Cooling Water Train A. Train A was aligned in standby while Auxiliary Component Cooling Water Train B equipment was out of service for scheduled maintenance. The system alignment was reviewed using Operations Procedure OP-002-001, Auxiliary Component Cooling Water, Revision 12.
  • Low-Pressure Safety Injection Train A: On February 25, 2002, the inspectors walked down the mechanical and electrical components of a critical portion of Low-Pressure Safety Injection Train A while Train A was in standby alignment.

This walkdown was completed while the Low-Pressure Safety Injection Train B pump was out of service for scheduled motor replacement. The review included the Updated Final Safety Analysis Report and Operations Procedure OP-009-008, Safety Injection System, Revision 16.

  • Auxiliary Component Cooling Water System: Over the period from March 8-22, 2001, the inspectors conducted a complete system walkdown of the auxiliary component cooling water system to verify proper mechanical and electrical alignment, adequate material condition, and labeling of associated components. The inspectors used Operations Procedure OP-002-003, Component Cooling Water System, Revision 13, and Piping and Instrumentation Drawing G-160 (sheets 1 through 6) to verify proper system alignment.

b. Findings No findings of significance were identified.

1R05 Fire Protection (71111.05)

a. Inspection Scope The inspectors conducted tours and assessed the material condition of the active and manual fire detection and suppression systems and ensured that combustible materials were appropriately controlled in the following areas:

  • Safeguards Pump Rooms A and B -35-foot elevation of the reactor auxiliary building on February 12, 2002
  • Emergency Feedwater Pumps A, B, and AB pump rooms/areas -35-foot elevation of the reactor auxiliary building on February 12, 2002
  • Component Cooling Water Heat Exchanger Rooms A and B +21-foot elevation on March 13, 2002
  • Turbine generator +67-foot elevation on March 22, 2002

-3-

  • Reactor auxiliary building +46-foot elevation on March 22, 2002
  • Reactor containment building -4-foot, +21-foot, and +46-foot elevations on March 24 and 25, 2002 b. Findings No findings of significance were identified.

1R07 Heat Sink Performance (71111.07)

a. Inspection Scope On January 31, 2002, the inspectors completed a review of the thermal performance test conducted on Wet Cooling Tower B in December 2001. The test was conducted in accordance with Procedure PE-004-033, Wet Cooling Tower Thermal Performance Test, Revision 0. The inspectors discussed the results of this test with the engineers responsible for data collection, interpretation, and trending. The inspectors reviewed the Updated Final Safety Analysis Report and the system description document.

b. Findings No findings of significance were identified.

1R11 Licensed Operator Requalification (71111.11)

a. Inspection Scope On February 4, 2002, the inspectors observed licensed operator requalification activities being conducted in the simulated control room. The simulator scenarios were part of the regularly scheduled licensed operator requalification cycle. The inspectors observed portions of the postexercise discussions conducted to determine if all critical objectives had been met.

b. Findings No findings of significance were identified.

1R12 Maintenance Rule Implementation (71111.12)

a. Inspection Scope The inspectors reviewed the Maintenance Rule data for the following to determine if the Maintenance Rule scope for these systems had been appropriate and reviewed the functional failure determinations and condition reports for the previous 18 months. In addition, the inspectors interviewed the Maintenance Rule coordinator:

-4-

  • Process Radiation Monitoring System: During the week of February 11, 2002, the inspectors completed a review of the process radiation monitoring system to ensure that the requirements of the Maintenance Rule were met. Several process radiation monitors had experienced problems in the recent past.
  • Nitrogen Gas System: During the week of February 19, 2002, the inspectors completed a review of the application of the Maintenance Rule to the nitrogen gas system. This system was classified as a(1) in accordance with the requirements of the Maintenance Rule due to exceeding the established performance criteria.
  • Annunciator System: During the week of February 25, 2002, the inspectors completed a review of the annunciator system to ensure that the requirements of the Maintenance Rule were met. Three functional failures had been experienced on this system in the last 18 months.
  • Feedwater Heater Drain System: During the week of March 4, 2002, the inspectors completed a review of the Maintenance Rule application to the feedwater heater drain system. During this inspection period, two plant transients were experienced due to failed feedwater heater level control valves.
  • Chemical and Volume Control System: During the week of March 4, 2002, the inspectors completed a review of the chemical and volume control system with regard to the application of Maintenance Rule requirements. The inspectors interviewed the assigned system engineer to determine if identified conditions had been properly screened for potential functional failures.

b. Findings No findings of significance were identified.

1R13 Maintenance Risk Assessments and Emergent Work Evaluation (71111.13)

.1 Auxiliary Component Cooling Water Pump B a. Inspection Scope On December 30, 2001, the Auxiliary Component Cooling Water Pump B outboard shaft sleeve bearing failed during a weekly chemical mixing of Wet Cooling Tower Basin Train B. The inspectors reviewed the licensees troubleshooting methods and results to verify that the emergent work procedures maintained plant conditions in a safe manner and that the equipment was properly restored after the bearing replacement. In addition, the inspectors reviewed Condition Report 2001-1399 and Maintenance Action

-5-Item 432644 and interviewed responsible engineers and operators to verify that appropriate consideration was given to the risks associated with the maintenance activities.

b. Findings No findings of significance were identified.

.2 Reactor Trip Breaker 7 a. Inspection Scope On January 14-15, 2002, the inspectors observed portions of the scheduled maintenance on Reactor Trip Breaker 7. The inspectors also reviewed the work controls associated with removing and reinstalling the breaker during online power operations and the equipment out-of-service risk model used to quantify the equipment outage risk to plant operations. The inspectors reviewed Maintenance Procedure ME-004-155, Reactor Trip Switchgear, Revision 12; Operations Procedure OP-903-127, Reactor Trip Circuit Breaker Post-Maintenance Retest, Revision 2; and Maintenance Action Item 429670.

b. Findings No findings of significance were identified.

.3 Motor Control Center 315B Planned Maintenance Outage a. Inspection Scope On January 30, 2002, the inspectors completed a review of the scheduled work for Motor Control Center 315B. The inspectors assessed whether the licensee had appropriately considered the risk of the scheduled work and that the licensees specified management controls were implemented. The inspectors reviewed Condition Report 2002-0154 and interviewed engineering, electrical maintenance, and work control center personnel responsible for work controls and job completion. In addition, the inspectors reviewed Maintenance Procedure ME-004-151, 480 Volt Motor Control Center (MCC), Revision 9. Motor Control Center 315B provided power and control to Ultimate Heat Sink Train B wet and dry cooling tower fans.

b. Findings No findings of significance were identified.

-6-

.4 Lifting and Handling Equipment a. Inspection Scope On January 30, 2002, the inspectors completed a risk assessment governing the work and work control process used to control the licensees lifting and handling program.

This inspection was begun as a result of lifting irregularities reported in the spent fuel pool. The inspectors reviewed Administrative Procedures MM-001-006, Sling Inspection and Control, Revision 5; MM-007-002, Crane and Hoist Inspection and Testing, Revision 5; and Condition Reports 2002-0083 and -0084. The inspectors interviewed engineering and construction personnel involved with the sites lifting and handling equipment.

b. Findings No findings of significance were identified.

.5 Charging Pump A a. Inspection Scope On January 31, 2002, the inspectors completed a review of emergent work performed on Charging Pump A. This pump was identified as having excessive water in the crankcase oil caused by pump seal leakage. The oil and filter were changed and the seals were replaced. The inspectors reviewed Maintenance Action Items 432565, 432969, 433100, and 433180 and interviewed plant operations personnel.

b. Findings The inspectors identified a violation of Technical Specification 6.8.1 for the failure to follow Procedure UNT-005-036, Reactivity Management Program, Revision 0, and Corporate Procedure WM-102, MAI Planning, Implementation, and Closeout, Revision 0. On January 18, 2002, the licensee performed emergent work to change the oil and oil filter in Charging Pump A, but failed to meet the requirements of the reactivity management program procedure. The finding was determined to affect the mitigating system cornerstone and to be of very low safety significance (Green) using the significance determination process.

The licensee determined that the oil in Charging Pump A was contaminated with water and needed to be changed along with the oil filter. This constituted emergent work since it did not go through the scheduled work planning process. While preparing the work package, the licensee identified that a reactivity management screening checklist was required, but failed to complete it as required by the applicable procedure.

The reactivity management program emphasizes the importance of a conservative operating policy such that the reactivity control of nuclear fuel, both in storage and in the reactor, is maintained in a safe and effective manner. This is, in part, achieved by

-7-screening maintenance activities for the possibility of an inadvertent reactivity change.

Step 5.2.2 of the procedure requires that each work package identified to impact reactivity management will have a review performed.

The inspectors were concerned that the reactivity management program was not being applied correctly to emergent work items such that potential inadvertent reactivity changes could take place during the performance of these emergent work activities.

Subsequent to this finding, the licensee performed a self-assessment to determine the extent of this situation. The licensee generated Condition Reports 2002-0476 and-0487, which detailed additional issues with the reactivity management program requirements. The inspectors considered these issues to be programmatic in nature in that the program requirements were not being met in all cases for maintenance work activities, particularly emergent work activities.

This violation was more than minor because it could be reasonably viewed as a precursor to a more significant event due to the potential for an unplanned reactivity excursion and could affect the function of the charging pump or other reactivity management systems. This issue was determined to be of very low safety significance because there was no inadvertent reactivity change.

The inspectors determined that the failure to meet the requirements of the reactivity management program procedure during maintenance activities constituted a violation of Technical Specification 6.8.1 and licensee Corporate Procedure WM-102, MAI Planning, Implementation, and Closeout, Revision 0, which require that written procedures be implemented. This violation is being treated as a noncited violation, consistent with Section VI.A.1 of the NRC Enforcement Policy and is in the licensees corrective program as Condition Reports 2002-0169 and -0476 (NCV 50-382/01009-01).

.6 Reactor Trip Circuit Breaker a. Inspection Scope On March 26, 2002, the inspectors completed a review of the work performed to correct a faulty reactor trip circuit breaker. The inspectors reviewed Operations Procedure OP-903-127, Reactor Trip Circuit Breaker Post-Maintenance Retest, Revision 3, Maintenance Action Item 430840, Condition Report 2002-0382, and the root cause analysis report prepared for the licensees Corrective Action Review Board.

b. Findings The inspectors identified a violation of Technical Specification 6.8.1 for the failure to follow Corporate Procedure WM-102, MAI Planning, Implementation, and Closeout, Revision 0. The finding was determined to affect the mitigating system cornerstone and to be of very low safety significance (Green) using the significance determination process. On March 6, 2002, the licensee performed corrective maintenance on a reactor trip circuit breaker and placed the breaker in service without documenting the troubleshooting or repair activities.

-8-The licensee was preparing to perform preventive maintenance on Reactor Trip Circuit Breaker 6. To accomplish this, the breaker would be removed, another reactor trip circuit breaker installed in its place, and Reactor Trip Circuit Breaker 6 would be taken to the electrical shop for maintenance. During the installation of the spare reactor trip circuit breaker, the breaker unexpectedly closed as it was being placed into service. As a result, the breaker was removed from the cubicle and the normal breaker was reinstalled and tested. The breaker was returned to the electrical shop.

The following day, a licensee electrician was assigned to check on the problem. He determined that a loose stationary contact on the 52 X-Contactor resulted in actuation of the relay when touched during the racking-in process. The position of the 52 X-Contactor was in close proximity to the racking handle. The electrician removed the contactor and replaced it with one taken from another reactor trip circuit breaker that was no longer used in the plant.

Later that day, the repaired reactor trip circuit breaker was installed in the plant in place of Reactor Trip Circuit Breaker 6 to allow performance of the originally scheduled preventive maintenance on that component.

The inspectors requested to see the documentation for the troubleshooting and repair of the spare breaker along with the emergent work form. No such documentation was available. The troubleshooting and repair of the breaker had been performed outside of the licensees established process for performing work on safety-related components.

The specific problem and the details of the repair parts used were not documented to allow independent verification of the maintenance activities by others outside of the electrical shop and to provide for traceability.

The safety significance of this violation was determined to be more than minor because there was a credible impact on safety by not performing corrective maintenance in accordance with established procedures on safety-related equipment (reactor trip circuit breaker), which could affect the operability, availability, reliability, or function of the reactor protection system. Using the reactor safety significance determination process, the violation was determined to have very low safety significance because the reactor trip breakers would have functioned if required.

The inspectors determined that the failure to perform corrective maintenance on safety-related equipment in accordance with established procedures constituted a violation of Technical Specification 6.8.1 and licensee Corporate Procedure WM-102, MAI Planning, Implementation, and Closeout, Revision 0, which specifies how such work is to be performed and documented. This is being treated as a noncited violation, consistent with Section VI.A.1 of the NRC Enforcement Policy and is in the licensees corrective action program as Condition Report 2002-0382 (NCV 50-382/01009-02).

-9-1R15 Operability Evaluations (71111.15)

a. Inspection Scope The inspectors reviewed the operability evaluations for the following:

  • Offsite Power Distribution Systems Affect on the Operability of Class 1E Electrical Distribution Train: On January 26, 2002, the inspectors completed a review of an operability evaluation for Condition Report 2001-1339. This operability evaluation considered whether the offsite power distribution system would supply sufficient power to maintain vital load centers following a trip of the main generator and reactor at Waterford 3. The inspectors considered whether the present grid stability design assumptions are still valid at the Waterford 3 facility. The inspectors reviewed Condition Report 2000-0581; the licensees response to a World Association of Nuclear Operators Significant Operational Event Report 99-01, Loss of Grid, NRC Information Notice 2000-06, Offsite Power Voltage Inadequacies; and OP-903-066, Electrical Breaker Alignment Check, Revision 7. The operability evaluation sought to determine if all significant aspects of the loss of offsite power were addressed as described in the Updated Final Safety Analysis Report.
  • Broad Range Gas Monitors A and B Operability: On February 15, 2002, the inspectors completed an evaluation of an operability assessment completed by the licensees engineering organization concerning the operability of Broad Range Gas Monitors A and B. The inspectors reviewed the written assessment and Condition Reports 2002-0230, -0232, and -0261 and interviewed the supervisor of electrical and instrument and control engineering. The communication link between the plant monitoring computer and the broad range gas monitors exhibited inconsistencies.
  • Reactor Trip Circuit Breaker Response Time Testing Procedure: On February 26, 2002, the inspectors performed a review of the root cause analysis report that was conducted in response to Condition Report 2002-0200. The condition described in the report involved a reactor trip circuit breaker response time testing procedure that resulted in rendering all four channels inoperable.

This was caused by the installation of jumpers that bypassed the actuation signals for the undervoltage trip functions for all eight breakers at the same time.

  • Back-leakage Through Check Valve SI-142A: On March 8, 2002, the inspectors completed a review of a confirmation of operability evaluation for a condition involving back-leakage through Check Valve SI-142A inside the containment isolation valve for Penetration 39, and Valve SI-138A, low-pressure safety injection header to Reactor Coolant System Loop 2B flow control valve. During performance of a high-pressure safety injection system test, the low-pressure safety injection became pressurized due to this back-leakage. The confirmation of operability evaluation was performed to determine if these systems were capable of performing their safety functions. The inspectors also reviewed Condition Report 2002-0353.

-10-b. Findings No findings of significance were identified.

1R16 Operator Workarounds (71111.16)

a. Inspection Scope On February 13, 2002, the inspectors reviewed the current operator workaround list and evaluated the effects on the operators abilities to implement the required actions during routine and accident conditions. The inspectors also reviewed the cumulative effects of long-standing operator workarounds and the potential for causing system misoperation, degrading mitigation system capabilities, and the impact on the operators responses to plant transients. The inspectors reviewed Procedure OI-002-000, Annunciator, Control Room Instrumentation and Workarounds Status Control, Revision 18.

b. Findings No findings of significance were identified.

1R17 Permanent Plant Modifications (71111.17)

a. Inspection Scope The inspectors reviewed procedures governing plant modifications to evaluate the effectiveness of the programs for implementing modifications to risk-significant systems, structures, and components, such that these changes did not adversely affect the design and licensing basis of the facility. The inspectors also reviewed 13 permanent plant modification packages and associated documentation, such as 10 CFR 50.59 review screens and safety evaluations, to verify that they were performed in accordance with plant procedures.

The inspectors interviewed the cognizant design and system engineers for the identified modifications as to their understanding of the modification packages.

The inspectors evaluated the effectiveness of the licensees corrective action process to identify and correct problems concerning the performance of permanent plant modifications. In this effort, the inspectors reviewed five corrective action documents and the subsequent corrective actions pertaining to licensee-identified problems and errors in the performance of permanent plant modifications.

b. Findings A noncited violation was identified for a design change (ER-W3-1999-0726-01-00) that failed to fully consider the requirements of Article NC-7000, Protection Against Overpressure, of Section III in the ASME Boiler and Pressure Vessel Code,

-11-1971 Edition through Winter 1972 Addenda. This failure resulted in the approval to install a relief valve with a setpoint greater than the design pressure in a section of pipe in a containment penetration that is normally isolated with entrained fluid.

In response to Generic Letter 96-06, Assurance of Equipment Operability and Containment Integrity During Design-basis Accident Conditions, the licensee developed design changes for six penetrations that would be isolated, with water entrained, during design-basis accident conditions. The team noted that five of the six penetrations were not isolated during normal plant conditions. The remaining penetration was normally isolated with entrained water.

The team found that the design engineers failed to consider the normally isolated condition and the effects of a normal plant heatup. As a result, the pressure in the isolated penetration piping could exceed the design pressure of 1950 psig. The team noted that the design engineers failed to recognize that the generic letter addressed both accident and normal conditions. As a result, licensee engineers specified a relief valve with a set pressure in excess of the design pressure (i.e., 2500 psig).

Criterion III of Appendix B to 10 CFR Part 50 requires, in part, that [m]easures shall be established to assure that regulatory requirements . . . are correctly translated into specifications . . . Criterion III further requires that [m]easures shall also be established for the selection . . . of materials, parts, [and] equipment . . . that are essential to the safety-related functions of the structures, systems, and components.

However, the measures established did not assure that the design basis was currently translated into specifications for isolated penetration piping. Specifically, design approvals were inappropriately received for certain penetration piping with relief valve setpoints in excess of the design pressure of the piping (NCV 50-382/01009-03).

This design change had a credible impact on safety because the design change directed the installation of a relief valve with a set pressure greater than the design pressure, as allowed by the ASME Code. This design change also could have affected the integrity of the containment barrier as a result of not providing overpressure protection such that the design pressure of any component within the boundary would not be exceeded.

The team then assessed this finding in accordance with the significance determination process. The team found that this finding affected containment integrity and resulted in a degraded condition. This issue was determined to be of very low safety significance because it did not represent: a degradation of the radiological barrier function provided for the control room, or auxiliary building, or spent fuel pool; a degradation of the barrier function of the control room against smoke or a toxic atmosphere; or an actual open pathway in the physical integrity of reactor containment or an actual reduction of the atmospheric pressure control function of the reactor containment.

The licensee entered this issue in the corrective action program by issuing Condition Report 2002-0079.

-12-1R19 Postmaintenance Testing (71111.19)

a. Inspection Scope The inspectors reviewed postmaintenance testing activities conducted on the following components:

  • Auxiliary Component Cooling Water Pump B: On January 22, 2001, the inspectors completed an evaluation of postmaintenance testing conducted on Auxiliary Component Cooling Water Pump B following an unplanned outage to repair the pumps outboard shaft bearing. The inspectors reviewed Condition Report 2001-1399; Maintenance Action Item 432644; and Operating Procedure OP-903-050, Component Cooling Water and Auxiliary Component Cooling Water Pump and Operability Test, Revision 16, along with the specified postmaintenance testing for each maintenance activity.
  • Component Cooling Water System Nitrogen Accumulator Replacement: On February 8, 2002, the inspectors completed a review of the postmaintenance testing conducted to verify the operation of Component Cooling Water System Valves CC-134B and -135B and their associated nitrogen accumulators. The original carbon steel accumulators were replaced with stainless steel accumulators as part of the licensees corrosion control program. The inspectors observed portions of the replacement activities and reviewed Maintenance Action Items 404096 and 426374. The postmaintenance testing was conducted using Surveillance Procedure OP-903-118, Primary Auxiliaries Quarterly IST Valve Tests, Revision 6.
  • Shield Building Ventilation A: During the week of February 18, 2002, the inspectors reviewed the Maintenance Action Item 433494 work package for a scheduled Shield Building Ventilation Train A maintenance outage. The inspectors reviewed the scope of the work performed and the postmaintenance testing requirements.
  • Essential Chiller Unit B: On March 14, 2002, the inspectors conducted a review of the postmaintenance testing conducted following corrective maintenance on Essential Chiller Unit B. This unit was operating when the guide vane linkage became disconnected from the guide vane actuator. This condition was repaired and tested. The inspectors reviewed Maintenance Action Item 434253, Condition Report 2002-0379, and the associated emergent work approval form.
  • Feedwater Heater Drain Valve FHD-455A: On March 22, 2002, the inspectors completed an evaluation of postmaintenance testing conducted on Feedwater Heater Drain Valve FHD-455A following an unplanned outage to repair the valve after the valve plug separated from the stem, resulting in an unexpected plant transient. The inspectors reviewed Condition Report 2002-0388 and Maintenance Action Item 434333 along with the specified postmaintenance testing for each maintenance activity.

-13-

  • Auxiliary Component Cooling Water System ACCEOL311A 10M Thermal Overload Relay: On March 29, 2002, the inspectors completed a review of the work scope and postmaintenance testing conducted to verify the operation of Component Cooling Water System ACCEOL311A 10M relay. The inspectors reviewed Maintenance Action Items 428516, 428517, 416032; Condition Report 2002-0541; Surveillance Procedure ME-003-330, 480 Volt G.E.

Switchgear Breakers, Revision 13; and the specified postmaintenance testing for the Auxiliary Component Cooling Water System ACCEOL311A 10M relay.

b. Findings No findings of significance were identified.

1R20 Refueling and Outage Activities (71111.20)

a. Inspection Scope The inspectors reviewed the Refueling Outage 11 outage plan and risk assessment to determine the level of risk and plant equipment available during the course of the outage. Portions of the plant shutdown and cooldown processes were observed. The inspectors monitored the licensees control of the first portion of outage activities including the following:

  • Monitored outage configuration management, including the activities of the operational risk assessment team that assessed the outage plan from a risk perspective
  • Reviewed reactor coolant system level instrumentation used during midloop operations to determine if it was installed and functioned as expected
  • Reviewed plant electrical lineups to determine if the designated protected train was maintained
  • Monitored shutdown cooling system operating parameters to establish if they were maintained within the required range
  • Reviewed reactor coolant system inventory control and reactivity control measures b. Findings No findings of significance were identified.

-14-1R22 Surveillance Testing (71111.22)

a. Inspection Scope The inspectors reviewed the following:

  • Component Cooling Water Makeup Pump B: On January 28, 2002, the inspectors observed a scheduled surveillance test on Component Cooling Water Makeup Pump B. The test was performed in accordance with Surveillance Procedure OP-903-129, Component Cooling Water Makeup Pump Operability Check, Revision 1.
  • Emergency Feedwater Pump A/B: On January 28, 2002, the inspectors observed a scheduled surveillance test and completed a review of the results of a quarterly inservice test of Emergency Feedwater Pump A/B. The test was performed in accordance with Surveillance Procedure OP-903-046, Emergency Feedwater Pump Operability Check, Revision 15. The purpose of the test was to verify operability of turbine-driven Emergency Feedwater Pump A/B, satisfy Technical Specification surveillance requirements, and stroke test the Emergency Feedwater Pump A/B turbine steam supply from Steam Generators 1 and 2.
  • Auxiliary Component Cooling Water Valve ACC-110B: The inspectors reviewed the valve operation test and evaluation system (VOTES) test conducted on March 4, 2002, for Auxiliary Component Cooling Water Valve ACC-110B. The inspection verified that the VOTES test met the required surveillance acceptance criteria. The test was conducted in accordance with Maintenance Action Item 416035 and Maintenance Procedures ME-007-048, VOTES Testing of Butterfly Valves, Revision 3, and ME-007-008, Motor-Operated Valves, Revision 12.
  • Auxiliary Component Cooling Water Pump B: The inspectors observed and reviewed the results of the inservice testing conducted on Auxiliary Component Cooling Water Pump B performed on March 7, 2002. The testing was conducted in accordance with Operations Procedure OP-903-050, Component Cooling Water and Auxiliary Component Cooling Water Pump and Valve Operability Test, Revision 16.
  • Reactor Fuel Receipt Inspection: The inspectors observed portions of the new fuel receipt inspection activities conducted in accordance with Procedure RF-002-001, Fuel Receipt, Revision 8, and Procedure UNT-008-031, New Fuel Receipt, Revision 2. This work included disassembling and reassembling the new fuel assembly shipping containers, unloading the new fuel assemblies from the shipping containers, inspecting the new fuel, lifting and handling the new fuel for inspection, and storing the new fuel in the spent fuel pool.

-15-

  • Main Steam Safety Valves MS-106A, -108A, and -110A: The inspectors observed and reviewed the functional inservice tests conducted on March 21, 2002, for Main Steam Safety Valves MS-106A, -108A and -110A. The inspectors observed the testing and reviewed the inservice surveillance test data and Maintenance Action Items 412392, 412390, and 412394. In addition, the inspectors reviewed Mechanical Maintenance Procedure MM-007-015, Main Steam Safety Valve Test, Revision 6.

b. Findings No findings of significance were identified.

1R23 Temporary Plant Modifications (71111.23)

a. Inspection Scope On March 13, 2002, the inspectors conducted a review of a temporary plant modification to install steam generator primary to secondary leakage detection while awaiting a permanent plant modification. The inspectors reviewed the following documents:

  • Engineering Request ER-W3-99-1118-00-04
  • Procedure UNT-005-032, Steam Generator Primary-to-Secondary Leakage, Revision 3
  • Procedure CE-003-705, Determination of Primary to Secondary Steam Generator Leak Rate, Revision 3
  • Procedure OP-003-001, Condenser Air Evacuation System, Revision 10
  • Procedure OP-901-202, Steam Generator Tube Leakage or High Activity, Revision 3 b. Findings No findings of significance were identified.

Emergency Preparedness 1EP6 Drill Evaluation (71114.06)

a. Inspection Scope The inspectors observed activities in the simulated control room, Emergency Operations Facility, Technical Support Center, and Operations Support Center, and reviewed the drill scenario. The drill scenario simulated equipment failures, a site evacuation, a reactor core transient, leakage of reactor coolant, and the release of radioactive material

-16-offsite. In addition, the inspectors reviewed the drill critiques and the resolution of identified performance weaknesses. The drill was conducted on February 19, 2002.

b. Findings No findings of significance were identified.

2 RADIATION SAFETY Occupational Radiation Safety 2OS2 ALARA Planning and Controls (71121.02)

a. Inspection Scope The inspectors interviewed radiation workers and radiation protection personnel and conducted independent radiation surveys and field observations for as low as reasonably achievable (ALARA) planning and controls of selected work areas within the controlled access area during normal operations. The inspectors attended a special ALARA committee meeting conducted February 28, 2002, to review and approve job dose estimates for radiation work packages that had estimated doses greater than 5 rem during Refueling Outage 11. In addition, the inspectors reviewed the following items and compared them with regulatory requirements to determine whether the licensee had an adequate program to maintain occupational exposures ALARA:

  • ALARA areas addressed in the Quality Assurance Audit QA-15-2001-W3-1, Radwaste, performed October 22 through November 13, 2001
  • Three quality assurance surveillances (QS-2001-W3-148, Pre-job Briefing of Short Cycle Recirculation/Fluff of the Spent Resin Tank, November 13, 2001; QS-2002-W3-009, Mechanical Seal Replacement on Fuel Pool Pump A, January 15, 2002; and QS-2002-W3-014, Pre-job Briefing for Containment Entry, January 24, 2002)
  • Processes used to estimate and track exposures
  • Plant collective exposure history for the past 3 years, current exposure trends, and 3-year rolling average dose information
  • Six radiation work permit (RWP) packages for Refueling Outage 11 work activities, which were anticipated to result in the highest personnel collective exposures during the refueling outage (RWP 2002-1508, RCP Motors (1A, 1B, 2A, & 2B) Work; RWP 2002-1511, Steam Generator #1 & #2 Primary Side Work (Perform Eddy Current & Tube Plugging); RWP 2002-1513, Steam Generator #1 Thermal Liner Repair; RWP 2002-1702, Reactor Head

-17-Disassembly; RWP 2000-1705, Reactor Head Re-assembly; and RWP 2002-1711, CEDM Nozzle Inspection on the Reactor Head

  • Use of engineering controls to achieve dose reductions, including six temporary shielding requests (2002-04, -10, -19, -25, -29, and -32) planned for installation during Refueling Outage 11
  • Individual exposures of selected work groups, including radiation protection, operations, mechanical maintenance, electrical maintenance, and instrument and controls maintenance
  • Hot spot tracking and reduction program
  • Radiological work planning
  • ALARA committee meeting minutes for the quarterly regular meeting conducted on December 10, 2001, and a special meeting conducted on August 22, 2001
  • Declared pregnant worker dose monitoring controls
  • A summary list of radiological worker performance and ALARA-related condition reports written since September 1, 2001 (11 condition reports from this list were reviewed in detail: CR-WF3-2001-1094, -1101, -1251, -1260, -1264, and -1366; and CR-WF3-2002-0029, -0176, -0294, -0295, and -0301)
  • Job site inspection and ALARA controls of work activities for the replacement of the Low-Pressure Safety Injection Pump B motor b. Findings No findings of significance were identified.

4 OTHER ACTIVITIES 4OA1 Performance Indicator Verification (71151)

a. Inspection Scope The inspectors reviewed initiating events cornerstone performance indicator data for the following:

  • Performance indicator data for unplanned scrams per 7,000 critical hours for the fourth quarter of 2001 on February 13, 2002
  • Performance indicator data for scrams with loss of normal heat removal for the fourth quarter of 2001 on February 26, 2002

-18-

  • Performance indicator data for unplanned power changes per 7,000 critical hours for the fourth quarter of 2001 on February 28, 2002 b. Findings No findings of significance were identified.

4OA3 Event Followup (71153)

a. Inspection Scope On March 23, 2002, an Alert condition was declared in response to plant operators inability to establish shutdown cooling conditions due to the failure of hydraulic-pneumatic valves in each train of the shutdown cooling water system.

Valves SI-405A and -405B failed to move to the open position when demanded signals were generated from the control room. The Alert was declared at 10:20 a.m. (CST) and exited at 11:50 a.m. after the valves were successfully opened. Shutdown cooling conditions were subsequently established and the plant was placed in Mode 5. The NRC Region IV Incident Response Center was activated in the monitoring mode and the Senior Resident Inspector responded to the site to monitor plant conditions. The inspector reviewed the licensees actions in the control room and in the technical support center during the Alert condition.

b. Findings No findings of significance were identified.

4OA6 Meetings Exit Meeting Summary

.1 The reactor inspectors presented the inspection results to Mr. E. Ewing, General Manager, Plant Operations, and other members of licensee management on January 18, 2002. Licensee management acknowledged the inspection findings.

The inspectors asked the licensee whether any materials examined during the inspection should be considered proprietary. While proprietary information was identified by the licensee, no proprietary information is contained in this report.

.2 The inspectors presented the inspection results to Mr. E. Ewing, General Manager, Plant Operations, and other members of licensee management at the conclusion of the inspection on March 1, 2002. The licensee acknowledged the findings presented.

The inspectors asked the licensee whether any materials examined during the inspection should be considered proprietary. No proprietary information was identified.

-19-

.3 The resident inspectors presented the inspection results to Mr. J. Venable, Vice President, Operations, and other members of licensee management at the conclusion of the inspection on April 3, 2002. The licensee acknowledged the findings presented.

The inspectors asked the licensee whether any materials examined during the inspection should be considered proprietary. No proprietary information was identified.

ATTACHMENT PARTIAL LIST OF PERSONS CONTACTED Licensee B. S. Allen, Director, Engineering S. Anders, Superintendent, Plant Security L. Borel, Senior Engineer, Licensing M. K. Brandon, Manager, Licensing L. Dauzat, Supervisor, Radiation Protection J. R. Douet, Manager, Operations E. C. Ewing, General Manager, Plant Operations R. M. Fili, Manager, Quality Assurance B. Fron, Superintendent, Plant Security C. Fugate, Manager, Technical Support T. Gaudet, Director, Planning and Scheduling B. Goldman, Outage ALARA Planner P. Gropp, Manager, Engineering A. Harris, Director, Nuclear Safety Assurance J. Herron, Vice President, Operations C. Lambert, Director, Engineering T. P. Lett, Superintendent, Radiation Protection D. Madere, Supervisor, Licensing D. Miller, ALARA Coordinator, Radiation Protection R. Osborne, Manager, System Engineering R. Peters, Acting Director, Nuclear Safety Assurance B. Pilutti, Supervisor, Radiation Protection J. Reese, Supervisor, Engineering J. A. Ridgel, Manager, Maintenance G. Scott, Licensing Engineer, Licensing P. Staunton, Supervisor, Engineering T. E. Tankersley, Manager, Training J. Venable, Vice President, Operations D. Viener, Supervisor, Engineering ITEMS OPENED, CLOSED, AND DISCUSSED Opened 50-382/01009-01 NCV Failure to meet the requirements of the reactivity management program procedure during maintenance work activities (Section 1R13.5).

50-382/01009-02 NCV Failure to perform corrective maintenance on safety-related equipment in accordance with established procedures (Section 1R13.6).

-2-50-382/01009-03 NCV Design control measures failed to prevent design and approval for installation of a relief valve with a set pressure in excess of the design pressure (Section 1R17b).

Closed 50-382/01009-01 NCV Failure to meet the requirements of the reactivity management program procedure during maintenance work activities (Section 1R13.5).

50-382/01009-02 NCV Failure to perform corrective maintenance on safety-related equipment in accordance with established procedures (Section 1R13.6).

50-382/01009-03 NCV Design control measures failed to prevent design and approval for installation of a relief valve with a set pressure in excess of the design pressure (Section 1R17b).

DOCUMENTS REVIEWED Safety Evaluations NUMBER DESCRIPTION REVISION 2001-014 Removal of EFW Flow Control Valves EFW- 0 223A, 223B, 224A, and 224B from the containment isolation valve table in the TRM 2001-017 ER-W3-98-0821-01-00, inactivation of a 0 portion of the primary water treatment plant 2001-021 ER-W3-99-0726-01-00, changes to 10 0 containment piping penetrations to ensure that thermally-induced overpressurization does not affect the integrity of the containment isolation system 2001-023 ER-W3-1999-0198-004, replace insulation on 0 top of reactor vessel

-3-2001-025 ER-W3-2000-0106-000, add a direct- 0 operated backpressure control valve in the bleedoff line from the reactor coolant pump seals to the volume control tank 2001-027 ER-W3-99-0184-09-00, add a branch 0 connection to the existing pressurizer surge line 2001-029 DCP-3521, reroute DCT sump pumps 6 discharge to CW system 2001-031 STP 432049, EDG A Voltage Regulator 0 Retest (special test)

2000-069 ER-W3-00-0890-00-00, MSIV design basis 0 Safety Evaluation Screenings NUMBER DESCRIPTION REVISION ER-W3-00-0597-00-00 Hairline crack on the conduit fitting 0 OP-100-014 Technical Specification and Technical 10 Requirements Compliance HP-001-107 High Radiation Area Access Control 14 EP-002-071 Site Protective Measures 17 MM-007-017 High Pressure Tilt Pad Bearing Inspection 3 NE-001-005 Preparation, Control, and Documentation of 4 Fuel Movement W4.201 Configuration Management 5

-4-ER-W3-00-0691-00-00 Redesign of ACCW Pump B Lube Oil Cooler 0 Cooling Water Supply Connection ER-W3-01-0430-00-00 Equivalency Evaluation for Westinghouse 0 HQC3010 Molded Case Circuit Breaker ER-W3-00-0350-01-01 Install Bullet Resistant Enclosure on the 0 Condensate Polisher Building OP-903-053 Fire Protection System 10 Condition Reports NUMBER 2001-0410 2001-0480 2001-1125 2001-1226 Calculations NUMBER DESCRIPTION REVISION EC-M97-004 Evaluation of Containment Penetration Piping in Response 2 to GL 96-06 Calculations:

NUMBER TITLE REVISION EC-I99-001 ESF Response Time Acceptance Criteria Basis 1

-5-Calculations:

NUMBER TITLE REVISION EC-I00-002 Main Steam Isolation Valve Nitrogen Dome Pressure A EC-M00-009 Closure Time Analysis for Main Steam Isolation 0 Valves MS-124 A & B EC-M97-062 Incorporated spring replacement for PSL-204 2 EC-M98-004 Design Basis Review for Main Steam Isolation 0 Valves MS-124 A & B Condition Reports:

CR-WF3-2001-0513 CR-WF3-2001-0251 CR-WF3-2001-0410 CR-WF3-2001-0099 CR-WF3-2001-0375 Design Changes:

ER-W3-1998-0642-02-07 ER-W3-1999-0726-01-00 ER-W3-2001-0127-00-00 ER-W3-1998-0888-00-01 ER-W3-1999-0851-00-01 ER-W3-2001-0200-00-00 ER-W3-1998-1025-00-00 ER-W3-2000-0839-00-00 ER-W3-2001-0399-00-00 ER-W3-1999-0550-00-02 ER-W3-2000-0890-00-00 ER-W3-2001-0404-00-00 ER-W3-1999-0551-00-02 ER-W3-2000-0926-00-00 Miscellaneous Documents:

NUMBER TITLE/SUBJECT REVISION/

DATE Notice of Enforcement Discretion for Entergy Station, May 1, Unit 3 (NOED 00-6-006) 2000

-6-Miscellaneous Documents:

NUMBER TITLE/SUBJECT REVISION/

DATE Waterford Steam Electric Station, Unit 3 - Issuance of July 6, Amendment re: Reduction in Operable Containment Fan 2000 Coolers in the Containment Cooling System (TAC MA6997)

SPEER 9301165 Equivalency evaluation of replacement disc for 10" WKM 0 saf-t-seal gate valves, CS-125A, B SQ-MN-059 Replacement of spring 0 W3F1-2000-0057 Request Enforcement Discretion April 27, 2000 W3F1-2000-0124 Amendment 165 to the Waterford 3 Operating License August 31, 2000 W3F1-2000-0125 Termination of TS 3.6.2.2 NOED to Allow Operation with August 31, One CFC Operable per Containment Cooling Train 2000 W3F1-2000-0133 Issuance of Amendment 165 October 6, 2000 Procedures:

NUMBER TITLE REVISION LI-101 10 CFR 50.59 Review Program 0, 1

-7-Procedures:

NUMBER TITLE REVISION MM-003-043 Containment spray isolation valve inspection and testing 2 OP-903-121 Safety systems quarterly IST valve test 4 STP 00408142 CCW makeup single failure modification acceptance test 0 Maintenance Action Items:

MAI 408142 MAI 408147 MAI 417356 MAI 424483 MAI 408143 MAI 413036 MAI 423011 MAI 427860 MAI 408146 MAI 417350 MAI 424186 LIST OF ACRONYMS USED ALARA as low as reasonably achievable ASME American Society of Mechanical Engineers CFR Code of Federal Regulations MAI maintenance action item NRC Nuclear Regulatory Commission RWP radiation work permit VOTES valve operation test and evaluation system

Retrieved from "https://kanterella.com/w/index.php?title=IR_05000382/2001009&oldid=2704693"