IR 05000346/2002019
| ML030310226 | |
| Person / Time | |
|---|---|
| Site: | Davis Besse  |
| Issue date: | 01/31/2003 |
| From: | Grobe J NRC/RGN-III |
| To: | Myers L FirstEnergy Nuclear Operating Co |
| References | |
| IR-02-019 | |
| Download: ML030310226 (33) | |
Text
ary 31, 2003
SUBJECT:
DAVIS-BESSE NUCLEAR POWER STATION NRC INTEGRATED INSPECTION REPORT 50-346/02-19
Dear Mr. Myers:
On December 31, 2002, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at your Davis-Besse Nuclear Power Station. The enclosed report documents the inspection findings which were discussed on January 15, 2003, with you and other members of your staff.
The inspection examined activities conducted under your license as they relate to safety and compliance with the Commissions rules and regulations and with the conditions of your license.
The inspectors reviewed selected procedures and records, observed activities, and interviewed personnel. For the entire inspection period, the Davis-Besse Nuclear Power Station was under the Inspection Manual Chapter 0350 Process. The Davis-Besse Oversight Panel assessed inspection findings and other performance data to determine the required level and focus of followup inspection activities and any other appropriate regulatory actions. Even though the Reactor Oversight Process had been suspended at the Davis-Besse Nuclear Power Station, it was used as guidance for inspection activities and to assess findings.
One finding of very low safety significance (Green) was identified in the report. This finding was determined to involve a violation of NRC requirements. However, because of the very low safety significance of the finding, and because it was entered into your corrective action program, the NRC is treating the issue as a Non-Cited Violation in accordance with Section VI.A.1 of the NRC s Enforcement Policy.
If you contest the subject or severity of the Non-Cited Violation, you should provide a response within 30 days of the date of this inspection report, with the basis for your denial, to the U.S. Nuclear Regulatory Commission, ATTN: Document Control Desk, Washington, DC 20555-0001, with a copy to the Regional Administrator, U.S. Nuclear Regulatory Commission - Region III, 801 Warrenville Road, Lisle, IL 60532-4351; the Director, Office of Enforcement, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001; and the Resident Inspector office at the Davis-Besse facility. Since the terrorist attacks on September 11, 2001, the NRC has issued two Orders (dated February 25, 2002, and January 7, 2003) and several threat advisories to licensees of commercial power reactors to strengthen licensee capabilities, improve security force readiness, and enhance access authorization. The NRC also issued Temporary Instruction 2515/148 on August 28, 2002, that provided guidance to inspectors to audit and inspect licensee implementation of the interim compensatory measures (ICMs) required by the February 25th Order. Phase 1 of TI 2515/148 was completed at all commercial nuclear power plants during calendar year (CY) 2002, and the remaining inspections are scheduled for completion in CY 2003. Additionally, table-top security drills were conducted at several licensees to evaluate the impact of expanded adversary characteristics and the ICMs on licensee protection and mitigative strategies. Information gained and discrepancies identified during the audits and drills were reviewed and dispositioned by the Office of Nuclear Security and Incident Response. For CY 2003, the NRC will continue to monitor overall safeguards and security controls, conduct inspections, and resume force-on-force exercises at selected power plants. Should threat conditions change, the NRC may issue additional Orders, advisories, and temporary instructions to ensure adequate safety is being maintained at all commercial power reactors.
In accordance with 10 CFR 2.790 of the NRC's "Rules of Practice," a copy of this letter and its enclosure will be available electronically for public inspection in the NRC Public Document Room or from the Publicly Available Records (PARS) component of NRC's document system (ADAMS). ADAMS is accessible from the NRC Web site at http://www.nrc.gov/reading-rm/adams.html (the Public Electronic Reading Room).
Sincerely,
/RA/
John A. Grobe, Chairman Davis-Besse Oversight Panel Docket No. 50-346 License No. NPF-3
Enclosure:
Inspection Report 50-346/02-19 See Attached Distribution
DOCUMENT NAME: C:\MyFiles\Copies\ML030310226.wpd To receive a copy of this document, indicate in the box:"C" = Copy without enclosure "E"= Copy with enclosure"N"= No copy OFFICE RIII RIII RIII RIII NAME Passehl/trn Clayton Lipa Grobe DATE 01/31/03 01/ /03 1/31/03 1/31/03 OFFICIAL RECORD COPY
REGION III==
Docket No: 50-346 License No: NPF-3 Report No: 50-346/02-19 Licensee: FirstEnergy Nuclear Operating Company Facility: Davis-Besse Nuclear Power Station Location: 5501 North State Route 2 Oak Harbor, OH 43449-9760 Dates: November 15 through December 31, 2002 Inspectors: S. Thomas, Senior Resident Inspector D. Simpkins, Resident Inspector R. Powell, Senior Resident Inspector (Perry Station)
M. Bielby, Senior Licensing Inspector J. Belanger, Senior Physical Security Inspector R. Kopriva, Senior Project Engineer (Region IV)
P. T. Young, Examiner J. House, Senior Radiation Protection Specialist Approved by: Christine A. Lipa, Chief Branch 4 Division of Reactor Projects
SUMMARY OF FINDINGS
IR 05000346-02-19, FirstEnergy Nuclear Operating Company, on 11/15-12/31/2002,
Davis-Besse Nuclear Power Station. Access Control to Radiologically Significant Areas.
This report covers a 6 week period of resident and baseline inspection. The inspection was conducted by resident, Region III, and Region IV inspectors. The significance of most findings is indicated by their color (Green, White, Yellow, Red) using Inspection Manual Chapter (IMC) 0609 Significance Determination Process (SDP). Findings for which the SDP does not apply may be Green or be assigned a severity level after NRC management review. The 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.
A. Inspector-Identified and Self-Revealing Findings
Cornerstone: Radiation Safety
- Green.
A finding of very low safety significance was identified through self revealing events. On two separate occasions, workers in containment received dose rate alarms on their electronic dosimeters and did not take the actions required by procedure DB-HP-01901, Radiation Work Permits Revision 7, and Radiation Work Permit (RWP) 2002-5571. These documents state that radiation worker response requirements for a dose rate alarm are to place the work in a safe condition, exit the work area, and notify Radiation Protection personnel of the alarm.
The finding was more than minor because if left uncorrected workers could receive a greater radiological exposure than was planned for, unnecessary exposure, and could lead to a performance indicator occurrence for unintended dose. The finding was of very low safety significance because the procedure violation was not an As Low As Is Reasonably Achievable issue, did not involve an overexposure, did not involve a substantial potential for an overexposure and did not compromise the licensees ability to assess dose. The finding was therefore
- Green.
The finding resulted from a violation of Technical Specification 6.8.1 which requires the implementation of radiation protection procedures. (Section 20S1.1)
B. Licensee Identified Findings No findings of significance were identified.
REPORT DETAILS
Summary of Plant Status
The plant was shutdown on February 16, 2002 for a refueling outage and to perform inspections of vessel head nozzles. During repair of one of the cracked control rod drive mechanism nozzles, significant degradation of the reactor vessel head was discovered. As a direct result of the need to resolve many issues surrounding the Davis-Besse reactor vessel head degradation, NRC management decided to implement Inspection Manual Chapter 0350, Oversight of Operating Reactor Facilities in a Shutdown Condition With Performance Problems. The fuel was removed from the reactor on June 26, 2002, and the plant remained shut down. For the entire inspection period, the Davis-Besse Nuclear Power Station was under the Inspection Manual Chapter 0350 Process. As part of this process, several additional team inspections continued. The subjects of these inspections included: Containment Health/Extent of Condition, System Health Assurance, Management and Human Performance, and Program Compliance. The results of these inspections will not be included as part of this inspection report, but upon completion, each will be documented in a separate inspection report which will be made publicly available on the NRC website.
REACTOR SAFETY
Cornerstones: Initiating Events, Mitigating Systems, and Barrier Integrity.
1R04 Equipment Alignment
a. Inspection Scope
The inspectors verified equipment alignment and identified any discrepancies that impacted the function of the system and potentially increased risk. The inspectors also verified that the licensee had properly identified and resolved any equipment alignment problems that would cause initiating events or impact the availability and functional capability of mitigating systems. Specific aspects of this inspection included reviewing plant procedures, drawings, and the Updated Safety Analysis Report (USAR), to determine the correct system lineup and evaluating any outstanding maintenance work requests on the system or any deficiencies that would affect the ability of the system to perform its function. A majority of the inspectors time was spent performing a walkdown inspection of the system. Key aspects of the walkdown inspection included verification that:
! valves were correctly positioned and did not exhibit leakage that would impact their function;
! electrical power was available as required;
! major system components were correctly labeled, lubricated, cooled, ventilated, etc;
! hangers and supports were correctly installed and functional;
! essential support systems were operational;
! ancillary equipment or debris did not interfere with system performance;
! tagging clearances were appropriate; and
! valves were locked as required by the licensees locked valve program.
During the walkdown, the inspectors also observed the material condition of the equipment to verify that there were no significant conditions not already in the licensees work control system. The inspectors performed a walkdown of the following systems:
! component cooling water; and
b. Findings
No findings of significance were identified.
1R05 Fire Protection
a. Inspection Scope
The inspectors conducted fire protection walkdowns which were focused on availability, accessibility, and the condition of fire fighting equipment, the control of transient combustibles, and on the condition and operating status of installed fire barriers. The inspectors selected fire areas for inspection based on their overall contribution to internal fire risk, as documented in the Individual Plant Examination of External Events (IPEEE),their potential to impact equipment which could initiate a plant transient, or their impact on the plants ability to respond to a security event. Using the documents listed at the end of this report, the inspectors verified that fire hoses and extinguishers were in their designated locations and available for immediate use, that fire detectors and sprinklers were unobstructed, that transient material loading was within the analyzed limits, and that fire doors, dampers, and penetration seals appeared to be in satisfactory condition.
The following areas or components were inspected:
! service water structure;
! emergency diesel generators; and
! containment fire loading evaluation.
b. Findings
No findings of significance were identified.
1R07 Heat Sink Performance
a. Inspection Scope
The inspectors reviewed the data from the latest performance test of decay heat exchanger 1-1. Through discussions with the engineer responsible for this heat exchanger and review of applicable documentation, the inspectors verified:
! the selected testing methodology was consistent with accepted industry practices;
! the test conditions were consistent with the selected methodology;
! the test acceptance criteria were consistent with the design basis values;
! the test results had appropriately considered differences between testing conditions and design conditions; and
! the frequency of the testing, based on trending data, was sufficient to detect degradation prior to the loss of heat removal capabilities below design basis values.
b. Findings
No findings of significance were identified.
1R11 Licensed Operator Requalification
.1 Facility Operating History
a. Inspection Scope
The inspectors reviewed the plants operating history from September 2001, through October 2002, to assess whether the Licensed Operator Requalification Training (LORT)program had addressed operator performance deficiencies noted at the plant.
b. Findings
No findings of significance were identified.
.2 Licensee Requalification Examinations
a. Inspection Scope
The inspectors performed a biennial inspection of the licensees LORT program. The inspectors reviewed the current year requalification biennial written examination and annual operating test material to evaluate general quality, construction, and difficulty level. The biennial written examination material consisted of 40 questions in a multiple-choice format. The questions addressed plant and control systems, administrative controls, and procedural limits. The operating test material consisted of dynamic simulator scenarios and job performance measures (JPMs). The inspectors reviewed the methodology for developing the examinations, including the LORT program 2 year sample plan, probabilistic risk assessment insights, previously identified operator performance deficiencies, and plant modifications. The inspectors assessed the level of examination material duplication during the current year annual examination (through four examinations). The inspectors also interviewed members of the licensees management and training staff, and discussed various aspects of the examination development.
b. Findings
No findings of significance were identified.
.3 Licensee Administration of Requalification Examinations
a. Inspection Scope
The inspectors observed administration of the requalification operating test to assess the licensees effectiveness in conducting the test and to assess the facility evaluators ability to determine adequate performance using objective, measurable performance standards. The inspectors evaluated, in parallel with the facility evaluators, the performance of five licensed operators for one operating shift crew during two dynamic simulator scenarios. The operating shift crew was divided into two simulator crews for evaluation purposes. Each simulator crew consisted of three Senior Reactor Operators and two Reactor Operators. The inspectors conducted reviews to verify that all licensed operators participated in at least two evaluated scenarios during the annual test or at some time during the annual training cycle. In addition, the inspectors observed licensee evaluators administer five JPMs to a select number of operators. The inspectors observed the training staff personnel administering the operating test, including pre-examination briefings, observations of operator performance, individual and crew evaluations after dynamic scenarios, techniques for JPM cuing, and the final evaluation briefing for licensed operators. The inspectors evaluated the adequacy of the simulator performance to support the examinations. The inspectors also reviewed the licensees overall examination security program.
b. Findings
No findings of significance were identified.
.4 Licensee Requalification Training Feedback Process
a. Inspection Scope
The inspectors assessed the effectiveness of the licensees processes for revision and maintenance of the LORT program, including the use of plant events and industry experience feedback information. The inspectors interviewed licensee personnel (operators, instructors, and management) and reviewed applicable procedures. In addition, the inspectors reviewed the licensees quality assurance and quality control oversight activities, including training and department self-assessment reports, to evaluate the licensees ability to assess effectiveness of the LORT program and implementation of appropriate corrective actions.
b. Findings
No findings of significance were identified.
.5 Licensee Remedial Training Program
a. Inspection Scope
The inspectors assessed the adequacy and effectiveness of remedial training administered to one individual that demonstrated unsatisfactory performance during an annual operating test scenario administered the previous week. The inspectors reviewed the training package to ensure that performance and knowledge weaknesses identified during the annual examination were adequately addressed. The inspectors also reviewed remedial training procedures and records to ensure that the subsequent re-evaluation was properly completed prior to returning the individual to licensed duties.
b. Findings
No findings of significance were identified.
.6 Conformance with Operator License Condition
a. Inspection Scope
The inspectors evaluated facility and individual operator license conformance with the requirements of 10 CFR Part 55. The inspectors reviewed the licensees program for maintaining active operator licenses to assess compliance with 10 CFR 55.53(e) and (f).
The inspectors reviewed the licensees procedural compliance and the process for tracking on-shift hours for licensed operators. The inspectors also conducted reviews to verify that proficiency watch-standing hours were credited to the correct control room positions in accordance with Technical Specifications. The inspectors reviewed six licensed operator medical records to ensure compliance with 10 CFR 55.21 and 55.25, and medical standards delineated in ANSI/ANS-3.4. In addition, the inspectors reviewed the licensees LORT program to assess compliance with the requalification program requirements prescribed by 10 CFR 55.59(c).
b. Findings
No findings of significance were identified.
.7 Written Examination and Operating Test Results
a. Inspection Scope
The inspectors reviewed the first 4 weeks pass/fail results of the 2002 annual written examinations and operating tests administered by the licensee and prescribed by 10 CFR 55.59(a)(2).
b. Findings
No findings of significance were identified.
.8 Conformance with Simulator Requirements
a. Inspection Scope
The inspectors evaluated conformance of the licensees simulation facility for use in administering the operating test, and as a plant-referenced simulator for satisfying experience requirements for applicants for license applications as prescribed in 10 CFR 55.46. The inspectors reviewed the licensees process for continued assurance of simulator fidelity with regard to identifying, reporting, correcting, and resolving simulator discrepancies. The inspectors reviewed simulator certification testing to assess compliance with standards delineated in ANSI/ANS-3.5, 10 CFR 55.46(c) and 55.46(d).
b. Findings
No findings of significance were identified.
.9 Simulator Requalification Observation
a. Inspection Scope
The inspectors observed an operating crew on the simulator during annual requalification examination activities. The inspectors observed two simulator scenarios ORQ-EPE-S113 and ORQ-EPE-S116. The inspectors evaluated crew performance in the areas of:
! clarity and formality of communications;
! ability to take timely actions in the safe direction;
! prioritization, interpretation, and verification of alarms;
! procedure use;
! control board manipulations;
! oversight and direction from supervisors; and
! group dynamics.
The inspectors also observed the performance of the examination evaluators, their critique of the crews performance, and the self-critique done by the operating crew to verify that any observed weaknesses were identified and documented by the licensee.
Additionally, the inspectors reviewed the simulator configuration compared to the actual control room to verify that they were as identical as practical.
b. Findings
No findings of significance were identified.
1R19 Post-Maintenance Testing
a. Inspection Scope
The inspectors reviewed post-maintenance testing activities associated with maintenance on important mitigating and support systems or components to ensure that the testing adequately verified system operability and functional capability with consideration of the actual maintenance performed. The inspectors used the appropriate sections of Technical Specifications and the USAR, as well as the documents listed at the end of this report, to evaluate the scope of the maintenance and verify that the work control documents required sufficient post-maintenance testing to adequately demonstrate that the maintenance was successful and that operability was restored. In addition, the inspectors reviewed CRs to verify that any minor deficiencies identified during these inspections were entered into the licensees corrective action system. The inspectors observed and evaluated test activities associated with the following:
! packing adjustment and packing loading check for DH-76;
! thrust check and limit switch adjustment, and packing loading check for CF-1A;
! thrust check and limit switch adjustment, and packing loading check for CF-1B;
! restoration of diesel fire pump fuel oil tank after fouling was discovered and corrected;
! electric fire pump seal replacement and retest; and
! station air compressor #2 testing, following vendor motor refurbishment.
b. Findings
No findings of significance were identified
1R22 Surveillance Testing
a. Inspection Scope
The inspectors witnessed the surveillance tests and test data to verify that the equipment tested met Technical Specifications, USAR, and licensee procedural requirements, and also demonstrated that the equipment was capable of performing its intended safety functions. The activities were selected based on its importance in verifying mitigating system capability. The inspectors used the documents listed at the end of this report to verify that the tests met the TS frequency requirements; that the tests were conducted in accordance with the procedures, including establishing the proper plant conditions and prerequisites; that the test acceptance criteria were met; and that the results of the tests were properly reviewed and recorded.
The following tests were observed and evaluated:
! emergency diesel generator #2 monthly run; and
! diesel fire pump monthly run.
b. Findings
No findings of significance were identified.
RADIATION SAFETY
Cornerstone: Occupational Radiation Safety (OS)
2OS1 Access Control to Radiologically Significant Areas (71121.01)
.1 Radiation Work Permit Review
a. Inspection Scope
The inspectors evaluated Condition Report 02-10075 and the associated corrective actions which documented radiation workers failing to follow procedure requirements in response to electronic dosimetry alarms while working in containment.
b. Findings
The inspectors identified one Green finding of very low safety significance, associated with a Non-Cited Violation that resulted from workers failing to follow procedure and radiation work permit requirements for responding to their electronic dosimeter dose rate alarms.
On December 8 and 10, 2002, two workers in containment received dose rate alarms on their electronic dosimeters and did not take the actions required by procedure DB-HP-01901, Radiation Work Permits Revision 7, and Radiation Work Permit 2002-5571. Radiation worker response requirements for a dose rate alarm are to place the work in a safe condition, exit the work area, and promptly notify radiation protection personnel of the alarm. These two examples illustrated the following weaknesses in the licensees radiological controls practices:
! workers failed to follow requirements of the RWP and site procedure DB-HP-01901, Radiation Work Permits, Revision 7;
! less than adequate communication of expectations by radiation protection personnel to the workers occurred regarding response to dosimeter alarms; and
! less than adequate assessment and implementation of job controls by radiation protection occurred to ensure the dosimeter alarms provided their intended purpose for protecting the workers.
The workers did not follow the requirements of a site procedure and the radiation work permit for the job.
The inspectors determined that failing to follow procedure and radiation work permit requirements related to dosimeter alarm response was a performance deficiency warranting a significance evaluation. The inspectors concluded that the finding was greater than minor in accordance with IMC 0612, Power Reactor Inspection Reports, Appendix B. This issue affected the occupational radiation safety cornerstone to ensure adequate protection of radiation workers from exposure to radioactive material and the attribute for programs and processes. Using the Occupational Radiation Safety Significance Determination Process, the procedure violation was not an As Low As Is Reasonably Achievable issue, did not involve an overexposure, did not involve a substantial potential for an overexposure and did not compromise the licensees ability to assess dose. Therefore, the finding is Green.
Technical Specification 6.8.1 requires, in part, that procedures be established, implemented and maintained that cover the activities recommended in Regulatory Guide 1.33, Appendix A, dated November 1972 which include procedures for radiation protection. Procedure DB-HP-01901, Radiation Work Permits Revision 7 (Section 4.3.3.c.1) requires, in part, that personnel are expected to respond to a dosimeter alarm by: reading the electronic dosimeter; placing plant equipment in a safe condition (if necessary); exiting the area; and contacting radiation protection. Contrary to this, on December 8 and 10, 2002, two individuals received dose rate alarms but failed to leave the area and contact radiation protection. The failure to follow a procedure requirement is a violation of Technical Specification 6.8.1. However, since the licensee documented this issue as Condition Report 02-10075 in its corrective action program, and because the violation is of very low safety significance, the violation is being treated as a Non-Cited Violation in accordance with Section VI.A.1 of the NRCs Enforcement Policy (NCV 50-346/02-19-02).
SAFEGUARDS
Cornerstone: Physical Protection
3PP4 Security Plan Changes (71130.04)
a. Inspection Scope
The inspectors reviewed Revision 21/Change 1 to the Davis Besse Nuclear Plant Security Plan to verify that the changes did not decrease the effectiveness of the submitted document. The referenced revision was submitted in accordance with the regulatory requirements of 10 CFR 50.54(p) by a licensee letter dated July 9, 2002.
b. Findings
No findings of significance were identified.
OTHER ACTIVITIES (OA)
4OA2 Routine Review of Identification and Resolution of Problems
.1 Licensee Resolution of Condition Reports Containing Mode Restraints
a. Inspection Scope
The inspectors began to review the licensees process of resolving issues that had been placed into their corrective action program and had also been assigned a restraint for resolution prior to entering a specific operational Mode. The inspectors obtained a listing, dated December 16, 2002, of open condition reports with assigned mode restraints. This list contained approximately:
! 11 Mode 1 restraints; 0 completed
! 57 Mode 2 restraints; 3 completed
! 212 Mode 3 restraints; 18 completed
! 1190 Mode 4 restraints; 39 completed
! 138 Mode 5 restraints; 8 completed, and
! 194 Mode 6 restraints; 64 completed.
Included as part of the corrective action to close out the condition reports that contained Mode restraints were attachments that specifically stated the corrective action taken to lift the Mode restraint. The inspectors evaluated a sampling of condition reports which contained completed corrective actions for restraints assigned to Mode 3, 4, 5 and 6.
b. Findings
No findings of significance were identified.
.2 Documentation of Inspection Finding Tracking Number
As documented in Inspection Report 50-346/02-17, Section 4OA2.2, the inspectors identified numerous examples of the improper implementation of the licensees corrective action program.
This finding was inadvertently not assigned a tracking number in IR 50-346/02-17. This deficiency will be corrected by assigning this Finding the number 50-346/02-17-03.
4OA3 Event Follow-up
.1 (Closed) LER 50-346/2002-006: Emergency Diesel Generator Exhaust Piping Not
Adequately Protected From Potential Tornado-Generated Missiles On August 11, 2002, the licensee identified that the last 6 feet of the diesel exhaust piping is not protected from tornado-generated missiles. The licensees review also identified that an exterior door to a main steam line room was similarly inadequate in protecting the Main Steam Safety Valves. As a result of this condition, the licensee concluded that they were in a condition prohibited by Technical Specifications, in that the current licensing basis requires systems vital to safe shutdown be enclosed in Class I structures designed to withstand tornado-generated missiles. On September 6, 2002, the licensee entered TS 3.8.1.2 due to both EDGs being inoperable due to inadequate missile protection and TS 3.7.1.1 due to the Main Steam Safety Valves being inoperable for the same reason. This condition has apparently existed since original plant construction. The licensees apparent cause investigation was still in progress at the end of the inspection period as was the final safety significance determination. The inspectors considered this to be an Unresolved Item (URI)
(URI 50-346/02-19-01), pending completion of further engineering evaluation by the licensee.
.2 (Closed) LER 50-346/2002-005-00: Potential Clogging of the Emergency Sump Due to
Debris in Containment On December 11, 2002, the licensee issued a revision to this LER to provide additional information regarding the potential clogging of the emergency sump due to debris in containment. This revision superseded LER 50-346/2002-005-00 in its entirety.
LER 50-346/2002-005-01 will be reviewed and documented in a subsequent inspection report.
4OA5 Other Activities
One of the key building blocks in the licensees Return to Service Plan was the Management and Human Performance Excellence Plan. The purpose of this plan was to address the fact that management ineffectively implemented processes, and thus failed to detect and address plant problems as opportunities arose. The primary management contributors to this failure were grouped into the following areas:
! Nuclear Safety Culture;
! Management/Personnel Development;
! Standards and Decision-Making;
! Oversight and Assessments;
! Program/Corrective; and
! Action/Procedure Compliance.
The inspectors had the opportunity to observe the day to day progress that the licensee made toward completing Return to Service Plan activities. Almost every inspection activity performed by the resident inspectors touched upon one of those five areas.
Observations made by the resident inspectors were routinely discussed with the Davis-Besse Oversight Panel members and were used, in part, to gauge licensee efforts to improve their performance in these areas on a day-to-day basis.
The following issues were selected because they occurred throughout the reporting period and illustrated examples of ongoing weaknesses in engineering, operations, and maintenance with respect to Standards and Decision-Making, Oversight and Assessments; and Program/Corrective Action/Procedure Compliance or challenged the ability of the inspectors to assess the current overall status of licensee performance.
.1 Resident Inspector Observations Related to Restart Readiness
a. Poor Maintenance Practices During Repack of the Electric Fire Pump The electric fire pump packing material was being replaced under a maintenance work order. During a walkdown of the system, the inspectors noted the packing was leaking profusely, even though the pump had been isolated, and that an air trap in the electric fire pump test header was spraying water on nearby components. The inspectors also noted that the pump casing drain line was fouled which caused packing leakage from the pump to overflow onto the floor. When questioned by the inspectors, the SRO overseeing the maintenance activities explained that the test header had been pressurized by a system lineup required to secure the diesel fire pump, but that the air trap should not have been spraying. The inspectors further questioned why the test header drain line was not draining to the floor drain, even though the isolation valve was open, and were informed it was clogged. An Auxiliary Operator (AO) responded to assist the SRO and commented he had noted the spray from the test header earlier, but had not contacted the SRO because he felt the SRO was too busy with the diesel fire pump. Operations supervision later stated this was not an acceptable communications protocol, and the AO should have contacted either the control room or the SRO for resolution.
The inspectors observed that maintenance workers did not have a copy of the maintenance work order or the appropriate maintenance procedure to work on the electric fire pump packing upon arrival at the work site. Upon questioning, the workers responded they had been sent by their supervisor to stop the leakage, and had left in such a hurry that the procedure and work order were left behind. When informed by the inspectors of the lack of documentation, the SRO requested the workers retrieve it immediately and perform no work until they retrieved it. After obtaining the appropriate work documentation, the workers explained the packing had not yet been adjusted and that leakage was expected. They did not however, know why the drain line was fouled, and proceeded to clear it by rapping on the small copper line with a screwdriver. This same screwdriver was later used to clear the test line drain valve. The maintenance practices used to clear both drain lines were later deemed inappropriate by operations management. The inspectors further questioned why the pump packing was leaking if the pump had been isolated, and were informed the pump isolation valves had leaked for some time.
The last observation made by the inspectors was that the individual tasked with making the adjustment of the packing while the pump was operating was wearing a loose-fitting overshirt, the tails of which were dangling near the pump casing. Since the packing would be adjusted while the pump was operating, the inspectors encouraged the SRO to have the maintenance worker remove the loose outer clothing while working around rotating equipment.
Although none of the issues discussed in this example were of more that minor safety significance or rose to the level of violations of regulatory requirements, they clearly illustrated material deficiencies; a clogged drain line on the test header, a clogged casing drain, a leaking air trap on the test header, at least one leaking isolation valve on the electric fire pump, and poor maintenance practices; a lack of rigor in adhering to work orders, poor communications, and potentially unsafe working conditions. This issue was documented in the licensee corrective action program as Condition Report 02-10203 and the inspectors were informed by the Director of Maintenance that coaching sessions had been conducted with the maintenance workers involved.
b. Unauthorized Impairment of a Spent Fuel Pool Negative Pressure Area Door Several doors leading to the spent fuel pool area are required to be closed as part of the technical specification requirement for the operability of the Emergency Ventilation System (EVS). The purpose of the EVS was to maintain a negative pressure boundary for the spent fuel pool area. With this boundary not maintained, the EVS cannot maintain a negative pressure on the Spent Fuel Pool area and no nuclear fuel movement is allowed in the fuel handling building.
Maintenance activities required one of these doors to be blocked open to facilitate equipment movement into containment. Security personnel had discussions with the Shift Manager, and erroneously assumed permission was granted to block the door open. When the door was blocked open, weather concerns prompted a temporary plywood cover to be installed limiting airflow but yet allowing equipment passage. Later that shift, a fuel inspection team obtained permission from the Shift Manager and began moving fuel in the spent fuel pool. An operator making a tour discovered the door impairment and fuel movement was stopped.
Although this incident demonstrates a lack of communication and failure to follow procedures, the door impairment was less than the maximum allowed opening in the spent fuel pool negative pressure boundary. Investigations showed turnover discussions were general in nature, and personnel assumed other parts of the organization were tending to the details. Verbal communications were less than adequate, and pre-job briefs did not include adequate detail to allow the discrepancies to be found. Station procedures for door and boundary impairment were not followed. This issue was not more than minor because the requirements of Technical Specifications were not violated. This issue was documented in the licensee corrective action program as Condition Report 02-9770.
c. Incorrect Danger Tag Issue While performing a walkdown of the auxiliary boiler feedpump 2 to ensure that a safe work isolation had been established, an operator noticed the danger tag that had been hung on valve CW271, was labeled CC271. When the clearance was prepared, the clearance tag was labeled incorrectly as CC271, but was actually hung on the desired valve, CW271. Although this error was found before work had commenced, this illustrates a weakness in the attention to detail during the preparation, review, and performance of establishing the isolation.
Although this example illustrates multiple violations of NOPP-OP-1001, Clearance/Tagging Program, the issue was considered minor because no work was completed under the incorrect clearance. This issue was documented in the licensee corrective action program as Condition Report 02-09491.
d. Improper Credit of Proficiency Watch Hours for Licensed Operators The inspectors identified that the Training Department incorrectly credited hours for watch standing proficiency to both licensed operators standing parallel watches. In accordance with 10 CFR 55.53(e), licensed operators required to maintain active licenses must stand a minimum of seven 8-hour or five 12-hour watches per calendar quarter. Operators can stand parallel watches; however, credit can only be given to the individual that assumes the responsibility and performs the duties associated with the position for the entire watch.
The Training Department reviewed both the unit log and the licensed operator proficiency manual on a quarterly basis to verify that licensed operators stand the minimum number of hours to maintain active licenses. The inspectors identified two instances in which the process used by Training to document the watch hours incorrectly credited proficiency hours for both the individual standing the parallel watch and the individual signed into the unit log. However, in both cases the operators had a sufficient number of additional watch standing hours to meet the minimum number required to be in compliance with 10 CFR 55.53(e). The potential impact of incorrectly documenting the parallel watch standing hours was that an operator may not meet the minimum required proficiency hours to maintain an active license. Although the Training Department did not effectively execute this evolution, this was considered a minor administrative issue and was documented in the licensees corrective action program as CR 02-09370.
.2 Observations of Deep Drain Valve Maintenance
During this extended outage, the licensee performed preventative or corrective maintenance on 71 valves which required the reactor coolant system to be drained to a level approximately 10 inches above the reactor coolant system hot leg centerline and 3 valves that required the reactor coolant system to be drained to a level approximately 18 inches below the reactor coolant system hot centerline. The inspectors monitored the overall progress of this project and evaluated the work of several valves while in progress. These evaluations included:
! review of the work package;
! observing maintenance in progress;
! ensuring ALARA principles were practiced;
! determining if appropriate FME practices were utilized for jobs that were not actively being worked; and
! appropriate post maintenance tests were identified in the work package.
The inspectors did not identify any findings of significance during the conduct of this inspection.
.3 Completion of Appendix A to TI 2515/148, Rev 1
The inspector completed the pre-inspection audit for interim compensatory measures at nuclear power plants, dated September 13, 2002.
.4 Evaluation of the Status of the Licensee High Energy Line Break Reanalysis
The inspectors followed up licensee resolution for NRC Information Notice 2000-20, Potential Loss of Redundant Safety-Related Equipment Because of the Lack of High-Energy Line Break Barriers, as part of the Problem Identification and Resolution portion of Inspection Procedure 71111.06. This was evaluated as part of this procedure to assess the potential for flooding of risk significant equipment with high temperature steam or water.
The licensees evaluation of IN2000-20 identified that design basis documentation pertaining to steam line breaks in the turbine building was potentially incomplete. For example, steam impingement effects from a postulated break in the turbine building on risk-significant high and low voltage switchgear room doors and component cooling water system doors have not been evaluated against standard review plan criteria.
Additionally, the auxiliary feedwater pump and component cooling water pump room ventilation systems communicate with the turbine building. The licensee has not rigorously reviewed these ventilation system configurations against the standard review plan criteria. The standard review plan criteria was developed to ensure, among other things, that 10 CFR 50 Appendix A, General Design Criteria for Nuclear Power Plants, was met for the initial plant design. Because of this potential design basis vulnerability, the licensee performed a risk evaluation of the configurations to determine a time line for resolution. The increase in core damage frequency was 5E-7 which did not exceed the Regulatory Guide 1.174 (An Approach for using Probabilistic Risk Assessment in Risk-Informed Decisions On Plant-Specific Changes to the Licensing Basis) threshold for being risk-significant. The licensee had determined that a more detailed evaluation and review needed to be performed and set a time line to complete these reviews by December 4, 2001. Pending further review, this item is an Unresolved Item (URI 50-346/2001-011-01).
Interim Review and Findings On December 15, 2002, the inspectors reviewed condition report CR 01-2019, Initial Results of Investigation into NRC Information Notice 2000-20", and the licensees Calculation No. C-NSA-000.02-010 Revision 1, Turbine Building High Energy Line Break Evaluation. Based on the results of the evaluation, the licensee concluded that:
! All plant areas identified, with the exception of the CCW pump room and the AFW pump room, are not affected by the consequences of the postulated pipe breaks. The pipe breaks are sufficiently away from the target areas such that they are beyond the direct impact of pipe whip or jet impingement.
! The CCW pump room walls will be subjected to pipe whip load and the jet impingement load from a high energy line break. Some structural damage will result from the pipe rupture and the harsh environment created will enter the room. It was determined that the equipment required for the safe shutdown of the plant located in the CCW room would not be in the direct path of the pipe whip or jet impingement.
! The high energy line break in the area of the AFW pump room may cause impingement into the floor openings of the pump room. Due to the physical separation for the floor openings into the two AFW pump rooms, it would be unlikely that a break on one line would result in a jet impingement into both AFW rooms at the same time. Also, there is sufficient distance from the floor level at 585'-0' to the AFW pumps that a pipe rupture would not result in a direct impingement onto the AFW pumps. The slab may be subjected to a pipe whip load, but the load would not result in structural damage of the slab.
The licensee has concluded that 1) not knowing to what extent the jet impingement needs to be modeled; 2) the uncertainty of previous evaluations that may or may not have been performed; and 3) the low PSA model risk significance, all of the issues encompassed by the turbine building high energy line break evaluation need resolution but do not constitute an immediate reactor safety concern or an operability concern. The resolution of these issues is being tracked as a Plant Issue and Condition Report CR 01-2019 remains open to ensure that the issues continue to get the proper attention and resources applied toward resolution. Based on this conclusion, URI 50-346/2001-011-01 remains open.
.5 Documentation of Inspection Finding Tracking Number
As documented in Inspection Report 50-346/02-17, Section 4OA5.2, the inspectors observed a licensee employee warning two other licensee employees about the presence of NRC inspectors.
This finding was inadvertently not assigned a tracking number in IR 50-346/02-17. This will be corrected by assigning this Finding the number 50-346/02-17-02.
4OA6 Meetings
.1 Exit Meeting
The inspectors presented the inspection results to Mr. Fast, Plant Manager, and other members of licensee management on January 15, 2003. The licensee acknowledged the findings presented. No proprietary information was identified.
.2 Interim Exit Meetings
Interim exits were conducted for:
! Licensed Operator Requalification, 71111.11B, with Mr. M. Roder, Operations Manager, on November 15, 2002.
! Safeguards Inspection with Mr. M. Roder on November 26, 2002.
KEY POINTS OF CONTACT Licensee A. Bless, Licensing D. Bondy, Licensed Operator Requalification Training Lead G. Dunn, Outage Manager R. Fast, Plant Manager D. Gerren, Steam Generator Engineer J. Grabnar, Manager, Design Engineering D. Imlay, Superintendent, E&C Maintenance M. Marler, Manager, Nuclear Training P. McCloskey, Manager, Regulatory Affairs G. Melssen, Maintenance Rule Coordinator L. Meyers, Chief Operating Officer, FENOC W. Mugge, Manager, Nuclear Security R. Pell, Manager, Chemistry and Radiation Protection J. Powers, Director, Nuclear Engineering R. Rishel, PRA Specialist M. Roder, Manager, Plant Operations J. Rogers, Manager, Plant Engineering R. Schrauder, Director, Support Services A. Schumaker, Supervisor, Access Control (Acting)
A. Stallard, Operations Support Supervisor M. Stevens, Director, Work Management J. Vetter, Quality Assurance Supervisor G. Wolf, Senior Licensing Engineer LIST OF ITEMS OPENED CLOSED AND DISCUSSED Opened 50-346/02-19-01 URI Final Evaluation of Apparent Cause Evaluation for LER 50-346/2002-006-00. (Section 4OA3.1)50-346/02-19-02 NCV Failure to Respond to Dosimeter Alarms. (Section 2OS1)50-346/02-17-02 FIN Inappropriate Licensee Notification of NRC Inspector Activity.
(Section 4OA5.5)50-346/02-17-03 FIN Inadequate Implementation of the Corrective Action Process Which Led to Not Identifying a Potentially Reportable Issue.
(Section 4OA2.2)
Closed 50-346/2002-006 LER Emergency Diesel Generator Exhaust Piping Not Adequately Protected From Potential Tornado-Generated Missiles.
(Section 4OA3.1)50-346/2002-005-00 LER Potential Clogging of the Emergency Sump Due to Debris in Containment. (Section 4OA3.2)50-346/02-19-02 NCV Failure to Respond to Dosimeter Alarms. (Section 2OS1)50-346/02-17-02 FIN Inappropriate Licensee Notification of NRC Inspector Activity.
(Section 4OA5.5)50-346/02-17-03 FIN Inadequate Implementation of the Corrective Action Process Which Led to Not Identifying a Potentially Reportable Issue.
(Section 4OA2.2)
Discussed 50-346/2001-011-01 URI Design Basis Documentation Pertaining to Steam Line Breaks in the Turbine Building Was Potentially Incomplete.
(Section 4OA5.4)
LIST OF ACRONYMS USED ADAMS Agency-wide Document Access and Management System AFW Auxiliary Feedwater AO Auxiliary Operator ASME American Society of Mechanical Engineers CCW Component Cooling Water CFR Code of Federal Regulations CR Condition Report DHR Decay Heat Removal DRP Division of Reactor Projects DRS Division of Reactor Safety EDG Emergency Diesel Generator EOP Emergency Operating Procedure EVS Emergency Ventilation System FENOC FirstEnergy Nuclear Operating Company IMC Inspection Manual Chapter IR Inspection Report IPEEE Individual Plant Examination of External Events ISLOCA Inter-System Loss of Coolant Accident JPM Job Performance Measure LER Licensee Event Report LOCA Loss of Coolant Accident LORT Licensed Operator Requalification Training NCV Non-Cited Violation NRC United States Nuclear Regulatory Commission OHS Office of Homeland Security PARS Publically Available Records RO Reactor Operator RWP Radiation Work Permit SSC System, Structure or Component SDP Significance Determination Process SFP Spent Fuel Pool SM Shift Manager SP Surveillance Procedure SRO Senior Reactor Operator TS Technical Specifications URI Unresolved Item USAR Updated Safety Analysis Report LIST OF
DOCUMENTS REVIEWED
1R04 Equipment Alignment
M041A Piping and Instrumentation Diagram - Service Water Pumps Rev. 24
and Secondary Service Water System
M041B Primary Service Water System Rev. 54
M041C Service Water System for Containment Air Coolers Rev. 25
OS-020 Operations Schematic - Service Water Sheet 1 Rev. 56
OS-020 Operations Schematic - Service Water Sheet 2 Rev. 25
M036A Component Cooling Water System Rev. 24
M036B Component Cooling Water System Rev. 30
M036C Component Cooling Water System Rev. 25
OS-021 Operations Schematic - Component Cooling Water Sheet 1 Rev. 28
OS-021 Operations Schematic - Component Cooling Water Sheet 2 Rev. 21
OS-021 Operations Schematic - Component Cooling Water Sheet 3 Rev. 9
M033B Decay Heat Train 1 Rev. 39
M033C Decay Heat Train 2 Rev. 16
OS-004 Operations Schematic - Decay Heat System Sheet 1 Rev. 32
OS-004 Operations Schematic - Decay Heat System Sheet 2 Rev. 4
1R05 Fire Protection
Fire Protection General Floor Plan Intake Structure Rev. 9
A223F Fire Protection General Floor Plan 585'-0" Level Rev. 14
Fire Hazards Analysis Report
DB-FP-00007 Control of Transient Combustibles Rev. 01
DSO-91-00086 Intra-company Memorandum - Negation of TERMS 5/30/91
Commitment 014852 Required to Revise Transient
Combustible Program
NLD-91-07753 Negation of TERMS Commitment 7/3/91
M016A Station Fire Protection System Rev. 43
1R07 Heat Sink Performance
DB-PF-4703 Decay Heat Cooler Performance Test (dated 1/31/02) Rev. 03
USAR, Volume 7, Emergency Core Cooling System Rev. 22
Section 6.3
1R11 Licensed Operator Requalification
ANSI/ Medical Certification and Monitoring of Personnel Requiring
ANS-3.4-1983 Operator Licenses for Nuclear Power Plants
ANSI/ Nuclear Power Plant Simulator for Use In Operator Training
ANS-3.5-1998 and Examination
AR-02-TRAIN-01 Davis-Besse Nuclear Quality Assessment Report, 1/28-4/16/02
CR 02-00306 Protective Action Recommendation Procedure Issue, Protective
Action Recommendation Training Need Identified for SROs
CR 02-00468 No Training Review for Plant Modifications
CR 02-00478 Nuclear Operations Training Staff Levels
CR 02-00495 Modifications Not Being Provided To Training As Required
By Procedure
CR 02-00496 Improvements for Documentation of Modification Training
Tracking
CR 02-3260 Preliminary Notification of Event on Licensed Operator
Requalification Exams
Licensed Operator Proficiency Manual Rev. 7
Licensed Operator Requalification Exam Sample Plan
2001-2002
Licensed Operator Requalification Training Program Rev. 6
Training Plan;11/15/01
Licensed Operator Requalification Training Program Rev. 7
Training Plan; 10/15/02
Licensed Operator Requalification Training Schedule,
Cycles 01-01 through 01-05, and 02-01 through 02-04
NT-OT-07001 Licensed Operator Requalification Program Rev. 6
NT-OT-07002 Instant Senior Reactor Operator Training Program Rev. 5
NT-OT-07003 Senior Reactor Operator Training Program Rev. 4
NT-OT-07004 Reactor Operator Training Program Rev. 5
NT-OT-07012 Operations Supervisory Team Training Program Rev. 3
NT-OT-07013 Simulator Design Control Rev. 2
NT-OT-07014 Simulator Physical Fidelity Rev. 2
NT-OT-07015 Simulator Functional Fidelity Rev. 1
NT-OT-07016 Simulator Instructor Control Functions Rev. 1
NT-OT-07017 Shift Manager Training Program Rev. 3
One Individual Simulator Evaluation Remediation Plan;
11/8/02
Open Simulator Work Order Report; 10/25/02
ORQ-EPE-S113 EOP Simulator Evaluation-Loss of TPCW Hi Level Tank Rev. 7
Level, RCS Leak, Loss of CRD CCW Flow, Loss of All AC
ORQ-EPE-S120 EOP Simulator Evaluation-FW Conductivity, Non-Isolatable Rev. 7
Steam Leak
ORQ-EPE-S116 EOP Simulator Evaluation-Partial Loss of Instrument Rev. 6
Air/Reactor Trip/Post Trip Overcooling
ORQ-EPE-S124 EOP Simulator Evaluation-Reactor Startup, Loss of Seal Rev. 4
Return, Steam Leak
P-OPS-1 Written Examinations and Quizzes for Operations Training Rev. 5
Programs
P-OPS-3 Requalification Walkthrough Examination Rev. 5
P-OPS-4 Development and Conduct of Continuing Training Simulator Rev. 9
Evaluations
P-OPS-8 Operations Training Instructor Technical Qualification Rev. 4
Program
Q3/2002 Performance Indicator Data Summary Report
Regulatory Guide Medical Evaluation of Nuclear Power Plant Personnel Rev. 1
1.134 Requiring Operator Licenses
Regulatory Guide Nuclear Power Plant Simulator Facilities for Use In Operator Rev. 3
1.149 Training and License Examinations, 10/01
Selection of Six Licensed Operator Medical Records
2002 Licensed Operator Curriculum Review Committee
Meeting Minutes
2002 LORT Annual Operating Test JPMs
2002 LORT Annual Operating Test Scenarios for first
weeks (October 21 and 28; November 4 and 11, 2002)
2002 LORT Biennial RO and SRO Written Examinations
(first 2 weeks)
2002 LORT Training Attendance Sheets
G-OPS-2 Development and Maintenance of Operations Training Unit Rev. 2
Instructional Packages
Simulator Test TAB01; Manual Reactor Trip
Simulator Test TAB04; Simultaneous Trip of All Reactor
Coolant Pumps
Simulator Test N06; 60 Minutes Drift Test
OPS-JPM-102 Upgrade an Event and Perform Notifications Rev. 1
OPS-JPM-004 Control Room Evacuation, Reactor Operator Actions in the Rev. 0
Control Room
OPS-JPM-017 Recover from Letdown Isolation Rev. 0
OPS-JPM-088 Perform Attachment 1 of the Turbine Trip AB Rev. 0
OPS-JPM-048 Energizing the NNI-X Cabinets Rev. 1
OPS-JPM-043 Manual Operation of the Emergency Diesel Generator 1 Rev. 1
or 2 from EDG Room
1R19 Post-Maintenance Testing
Mechanical Packing Valves Rev. 07
Maintenance
Procedure
DB-MM-9059
Work Order DH76: Repack During 13 Refueling Outage Deep Drain Rev. 00
2-3620-000
Work Order CF1A: Repack, Replace Packing Gland Studs, Pins, and Nuts Rev.00
2-5687-000
Work Order Repack CF1B and Replace Packing Gland Studs, Pins, and Rev. 00
2-5596-00 Nuts
Work Order Disassemble CF1B as Required, Troubleshoot Cause of Stem Rev.00
2-5596-01 Score, Replace Valve Stem, and Reassemble Using a New
Work Order Remove Motor/Return to Vendor/ Reinstall
2-6431-004
DB-SS-04013 Station Air Compressor No. 2 Performance Check Rev. 02
DB-FP-04047 Diesel Fire Pump Test Rev. 01
DB-OP-06610 Station Fire Suppression Water System Rev. 03
Work Order Packing gland on pump outboard runs hotter than desired Rev. 04
2-7663-000
Work Order DFP speed slowly decreased Rev. 05
2-7717-000
CR 02-10222 Diesel Fire Pump Day Tank Contaminated
CR 02-10189 DFP Speed Decrease
Test Data Sheet for CF1A Unseating and Closing Thrust Values,
dated 12/06/02
Test Data Sheet for CF1B Unseating and Closing Thrust Values,
dated 12/12/02
1R22 Surveillance Testing
DB-SC-03071 Emergency Diesel Generator Monthly Test Rev. 03
DB-FP-04047 Diesel Fire Pump Test Rev. 01
2OS1 Access Control to Radiologically Significant Areas
DB-HP-01901 Radiation Work Permits Rev. 7
2002-10075 Radiation Work Permit, Replace Thermo-well RTD Bosses - Rev. 0
RCS East and West Hot Legs;
4OA2 Problem Identification and Resolution
MODE 6
CR 02-04336 CRNVS Equipment Requirements During Fuel Handling in
Modes 5 and 6.
CR 02-04752 Latent Issue Review - Emergency Diesel Generator - Fire
Damper FD1036 Possible Obstruction; Nuclear Operating
Administrative Procedure
CR 02-00794 Containment Purge Valve CV5007 Failed Stroke Time
CR 02-02903 Boric Acid on DH-136
CR 02-03022 Midland II Head Nozzle No. 64 Contract Variation 21352-9
Use-As-Is Disposition
CR 02-03114 Decay Heat Valve 14A
CR 02-03161 Thread Stripped on Manual Actuator of DH-14A
CR 02-03175 Tapped Hole on DH-14A Requires Repair
CR 02-03216 #1 Service Water Pump Motor Connection Box Has Missing
Screws
CR 02-03238 SW Pump #1 Strainer Handhole Cover Leak
CR 02-03337 Documentation Could Not Be Located
CR 02-03339 Reactor Cavity Seal Plate Seal Clamp
CR 02-03478 EDG #2 Room Temperature
CR 02-03508 RCM 5052 Low Flow Switch Failed to Actuate
CR 02-03542 Potential Non-Q Material Installed on Decay Heat Pump #2
Rotating Element
CR 02-03550 Operability Determination Concluded an SSC is Inoperable
CR 02-03654 Broken Insulator on Connection Post
CR 02-03660 Containment Purge Radiation Monitor 5052 Test Failure
CR 02-03662 CV-5003A Did Not Fully Close During Testing
CR 02-03711 LIR Review- EDG - Nuisance Alarm at Local EDG Panel for
Alternate Shutdown
CR 02-03833 Ineffective Implementation of Corrective Action For CR 01-2820
CR 02-03990 Failure of EDG1 Overspeed Trip Test
CR 02-04336 CRNVS Equipment Requirements During Fuel Handling in
Mode 5 and 6
CR 02-04390 SHRR/ EDG 1-2 Ventilation
CR 02-04561 LIR - EDG 2 Cabinet C3618 Raceway Cover Screw Missing
CR 02-04576 LIR - EDG 2 Generator Termination Cabinet Conduit Bushing
Loose
CR 02-04629 LIR - Emergency Diesel Generator 1-2 Fuel Oil System
CR 02-04752 LIR - EDG - Fire Dampner FD 1036 Possible Obstruction
CR 02-05049 PR/LMAP: Undocumented Sample Frequency Changes
CR 02-05110 FME in the Refuel Canal - Deep End
CR 02-05123 Issue with CCW Flow to Decay Heat Coolers - Based on
CR 02-03278 G.I. Review
CR 02-05340 Could not Recirc BAAT 1 Per Procedure
CR 02-05508 P42-2 Oil V-Rings Not Installed Correctly
CR 02-05584 Replacement Reactor Head
CR 02-06074 LIR: EDG Exhaust Piping Stress Problem Does Not Meet
Vendor Limits for Adapter
CR 02-06230 LIR EDG - Missing Minimum Wall Calculation in Calc. 123B/C4
CR 02-06240 LIR: EDG Fuel Oil Procurement Does Not Commitment Per
Log 950 LTR
CR 02-06288 #2 Decay Heat Pump Mechanical Seals Leaking
CR 02-06466 LIR: EDG Soakback Pump Equivalency
CR 02-06665 LIR - EDG The Operating Temperature of the Governor Actuator
is Not Known
CR 02-06882 LIR: EDG Lube Oil., Jacket Water & Generator Bearing Oil
Temperature
CR 02-06993 LIR - EDG Main Bearing Temperature Limits
CR 02-08010 LIR - EDG General Electric SBM Switches Failure (IN 98-19)
CR 02-08708 EVS Fan #1 Flexible Discharge Boot Leakage
MODE 5
CR 02-01062 Loose Fuel Rod in Fuel Assembly NJ100U
CR 02-01483 Foreign Material in Refueling Canal
CR 02-02042 Incomplete Dimension Recordings on Data Sheet
CR 02-02693 Inadequate VT-2 Qualification of Personnel
CR 02-04119 LIR-RCS: TE-RC-13-1 is not Contacting the RC13A Valve Body
CR 02-04120 LIR-RCS Walkdown: ID Tag Deficiencies
CR 02-04260 SHRR Main Steam Valve Packing Followers
CR 02-05491 LIR-SW: Bent/Damaged Instrument Tubing
MODE 4
CR 01-02803 ISI Examination of HPI Pump #2 Casing Studs
CR 02-00690 Leakage Detected During LLRT of Pen 102 Electrical Penetration
Assemblies
CR 02-00831 Turbine Control Valve Stem Seal Leakoff Line Damage
CR 02-00965 ICS-11AS, #2 Atmospheric Vent Valve Air Drop Test Exceeds 5%,
Per DB-PF-03440
CR 02-01138 Oil Found on Cold Leg Piping
CR 02-01166 OTSG OEM Plugged Tube Stabilization
CR 02-01403 Catastrophic Failure of Limit Switch Compartment Gasket
CR 02-05190 ORR - System Condition Report for Steam Generators
4OA3 Event Follow-up
LER Emergency Diesel Generator Exhaust Piping Not Adequately
2002-006 Protected From Potential Tornado-Generated Missiles
LER Potential Clogging of the Emergency Sump Due to Debris in
2002-005 Containment
Revision 00
4OA5 Other Activities
Work Order CF-30 - Open and Inspect to Determine Cause of the Banging Rev. 00
2-2983-00 and What Damage May Be Occurring.
Work Order Remove Bonnet and Internals for HP50 to Provide Access for Rev. 00
2-3355-00 the Inspection of the HPI Thermal Sleeve
Work Order Remove Bonnet and Internals for HP51 to provide Borescope Rev. 00
2-3356-00 Access for the Inspection of the HPI Thermal Sleeve
CR Fire Pump Issues Noted During Repacking of Electric Fire
2-10203 Pump
CR Electric Fire Pump Packing Gland Temperature
2-10051
Work Order Core Flood Tank 1 to Reactor Check - Thread Engagement on Rev. 04
2-6370-000 Body to Bonnet Nuts Insufficient
Work Order Core Flood Tank 2 to Reactor Check - Repack CF 28, W/O 02- Rev. 04
2-6361-000 5597-000
CR SFP Negative Pressure Area Door Impaired, Potential T.S.
2-09770 3.9.12 Violation
CR Incorrect Danger Tag Found on Valve
2-09491
Drawing Plant Elevation 623'-0' Rev. 11
M102
Drawing Plant at Elevation 603'-0' Rev. 17
M103
Drawing Plant at Elevation 585'-0' Rev. 12
M104
Drawing Plant at Elevation 566'-0' & 567'-0' Rev. 5
M105
Drawing Containment & Auxiliary Building Plan El. 623'-0' Rev. 15
M-121
Drawing Containment & Auxiliary Building Plan El. 603'-0' Rev. 17
M-122
Drawing Containment & Auxiliary Building Plan El. 585'-0' Rev. 27
M-123
Drawing Containment & Auxiliary Building Plan El. 565'-0' Rev. 18
M-124
30