IR 05000528/2002005
| ML020840013 | |
| Person / Time | |
|---|---|
| Site: | Palo Verde  |
| Issue date: | 03/20/2002 |
| From: | Gody A Operations Branch IV |
| To: | Overbeck G Arizona Public Service Co |
| References | |
| IR-02-005 | |
| Download: ML020840013 (19) | |
Text
March 20, 2002
SUBJECT:
PALO VERDE NUCLEAR GENERATING STATION, UNITS 1, 2, AND 3 - NRC INSPECTION REPORT 50-528/02-05; 50-529/02-05; 50-530/02-05
Dear Mr. Overbeck:
On February 1, 2002, the NRC completed an inspection at your Palo Verde Nuclear Generating Station, Units 1, 2, and 3. The enclosed report documents the inspection findings, which were discussed on February 1, 2002, with Mr. William Ide, Vice President, Nuclear Production, and other members of your staff and on March 19, 2002, with Mr. Michael Sontag.
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.
On the basis of the sample selected for review, there were no findings of significance identified during the inspection. The inspectors concluded that problems were properly identified, evaluated and resolved within the problem identification and resolution program.
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 NRCs document system (ADAMS). ADAMS is accessible from the NRC Web site at http://www.nrc.gov/reading-rm/adams.html (the Public Electronic Reading Room).
Sincerely,
/RA/
Anthony T. Gody, Chief, Operations Branch Division of Reactor Safety Dockets: 50-528; 50-529; 50-530 Licenses: NPF-41; NPF-51; NPF-74
Arizona Public Service Company-2-
Enclosure:
NRC Inspection Report 50-528/02-05; 50-529/02-05; 50-530/02-05
REGION IV==
Dockets:
50-528; 50-529; 50-530 Licenses:
NPF-41; NPF-51; NPF-74 Report No.:
50-528/02-05; 50-529/02-05; 50-530/02-05 Licensee:
Arizona Public Service Company Facility:
Palo Verde Nuclear Generating Station, Units 1, 2, and 3 Location:
5951 S. Wintersburg Road Tonopah, Arizona Dates:
January 21 through February 1, 2002 Inspectors:
G. Johnston, Senior Operations Engineer, Operations Branch M. Murphy, Senior Operations Engineer, Operations Branch H. Bundy, Senior Operations Engineer, Operations Branch G. Warnick, Resident Inspector, Projects Branch D Approved By:
A. Gody, Chief Operations Branch Division of Reactor Safety
SUMMARY OF FINDINGS IR 05000528-02-05; 05000529-02-05; 05000530-02-05, on 1/21-2/1/2002, Arizona Public Service Company. Palo Verde Nuclear Generating Station, Units 1, 2, and 3; biennial inspection of identification and resolution of problems.
The inspection was conducted by three regional senior operations engineers, and a resident inspector. The significance of most findings is indicated by their color (Green, White, Yellow, Red) using IMC 609, "Significance Determination Process." The NRC's 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 Significance Determination Process does not apply are indicated by "No Color" or by the severity level of the applicable violation.
Identification and Resolution of Problems The licensee was generally effective at identifying problems and placing them into the corrective action program. The licensee effectively used risk information in prioritizing the extent of evaluation of individual problems and the schedule for implementation of corrective actions. The licensee effectively prioritized and evaluated issues with few exceptions. One exception involved a final operability evaluation, which concluded that the main steam and feedwater isolation system actuation circuitry was operable, took approximately 5 months to complete. Another example involved a failure to fully determine the extent of a condition associated with Borg-Warner check valve failures, which resulted in additional failures.
Corrective actions, when specified, were implemented in a timely manner. Based on interviews conducted during this inspection, workers at the site felt free to input safety issues into the problem identification and resolution program (Section 4OA2).
Report Details 4OA2 Identification and Resolution of Problems a.
Effectiveness of Problem Identification (1)
Inspection Scope The team reviewed items selected across the seven safety cornerstones of safety to determine if problems were being properly identified, characterized, and entered into the corrective action program for evaluation and resolution. Specifically, the team selected approximately 70 condition reports/disposition requests, which had been issued between January 2001 and January 2002. The team also reviewed 2 licensee audits covering the corrective action program and 26 self-assessments of plant activities. One of the self assessments specifically addressed integrated issues resolution. The effectiveness of the audits and assessments were evaluated by comparing the audit and assessment results against self-revealing and NRC-identified issues.
The team evaluated the condition report/disposition requests to determine the licensees threshold for identifying problems and entering them into the corrective action program.
Also, the licensees efforts in establishing the scope of problems were evaluated by reviewing pertinent control room logs, radiation protection logs, work orders, audit and self-assessment results, action plans, and results from surveillance tests and preventive maintenance tasks. The condition report/disposition requests and other documents listed in the attachment to this report were used to facilitate the review.
(2)
Issues and Findings The team determined that the licensee was generally effective at identifying problems and entering them into the corrective action program. In general this was evidenced by the relatively few deficiencies identified by external organizations (including the NRC)
that had not been previously identified by the licensee during the review period. During this inspection most conditions adverse to quality were being handled within the corrective action program.
NRC Inspection Report 50-528/529/530-2001-003 documented an example where the licensee identified a failure to place a condition adverse to quality into the corrective action program during an investigation of Borg-Warner check valve failures. The issue involved failure to document a valve misalignment during a high pressure safety injection and check valve full flow test. Subsequent licensee actions adequately addressed this issue.
A number of operator error-related issues were identified during the period, which were appropriately placed into the corrective action program. A number of issues associated with control element assembly were identified during the period. The team found that the licensee effectively determined the scope and extent of the problem.
The team, observed that the licensees self-assessments in the areas of radiation protection and emergency planning were narrowly focused. The team noted that the two self assessments conducted in 2001 for radiation protection covered annual radiation protection decommissioning records and radioactive material control. There was insufficient breadth to cover overall program implementation. Other self-assessment areas were thoroughly reviewed.
The team found that audits by the nuclear assurance organization were effective in identifying areas of needed improvement and noncompliance with station procedures.
Findings and recommendations from those audits were placed in the corrective action program.
No findings of significance were identified.
b.
Prioritization and Evaluation of Issues (1)
Inspection Scope The team reviewed approximately 70 condition reports/disposition requests and supporting documentation, including an appropriate analysis of the cause of the problem, to ascertain whether the licensee's evaluation of the problems identified and considered the full extent of conditions, generic implications, common causes, and previous occurrences. In addition, the team reviewed the licensee's evaluation of selected industry experience information, including operating event reports and NRC and vendor generic notices, to assess if issues applicable to the Palo Verde Nuclear Generating Station were appropriately addressed. Specific items reviewed are listed in the attachment to this report.
(2)
Issues and Findings The team found that the licensee effectively prioritized and evaluated issues with few exceptions. The team noted that the licensee typically investigated issues with sufficient depth and breadth to determine both the scope and extent of condition. Most notable was the licensees effective review of control element assembly degradation, which was clear and thorough.
One notable exception was documented as a licensee-identified noncited violation in NRC Inspection Report 50-528/529/530-2001-003. The issue involved a failure to determine the extent of condition for Borg-Warner check valve failures that occurred in 1997 and 1998, and a subsequent failure on October 23, 2000. The team reviewed subsequent corrective actions and determined that they adequately addressed the issues sufficiently to prevent recurrence and were appropriately scheduled in accordance with the risk significance of the check valves in question.
Another notable exception involved the teams review of Licensee Event Report 50-529/2001-002-00 for Unit 2. The team noted that the resolution of design issues and a final operability evaluation, which concluded that the main
steam and feedwater isolation system actuation circuitry was operable took approximately 5 months to complete. Licensee Event Report 50-529/2001-002-00 described a main steam and feedwater isolation system logic board failure that resulted in a Unit 2 reactor trip on July 13, 2001. This issue is discussed in Section 4OA3 of this report.
With regard to determination of causal factors and root cause analysis the team did not identify any instances where the underlying causes were not appropriately categorized.
No findings of significance were identified.
c.
Effectiveness of Corrective Actions (1)
Inspection Scope The team reviewed condition reports/disposition requests, audits and self-assessments to verify that corrective actions related to the issues were identified and implemented in a timely manner commensurate with safety, including corrective actions to address common cause or generic concerns. The team also interviewed plant personnel to independently verify and assess the effectiveness of corrective actions implemented by the licensee. A listing of specific documents reviewed during the inspection is included as the attachment to this report.
(2)
Issues and Findings The licensee identified one example of ineffective corrective actions involving Borg-Warner check valve failures, which is discussed in Section 4OA2b(2) above. Based on a review of the licensees records, the team identified no further examples of ineffective licensee corrective actions.
No findings of significance were identified.
d.
Assessment of Safety Conscious Work Environment (1)
Inspection Scope The team interviewed 17 individuals from the licensee's staff, which represented a cross-section of functional organizations and supervisory and non-supervisory personnel. These interviews assessed whether conditions existed that would challenge the establishment of a safety conscious work environment.
(2)
Issues and Findings Based on interviews, the team identified no findings related to the safety conscious work environment. The team concluded, based on information from these interviews, that employees were willing to identify issues and accepted the responsibility to proactively identify and enter safety issues into the corrective action program.
No findings of significance were identified.
4OA3 Event Follow-up (71153)
(Open) Licensee Event Report 50-529/2001-002-00: Logic Board and Pin Connector Failure Causes Three of Four Main Steam Isolation Valves to Close. On July 13, 2001, a logic board failure in the main steam and feedwater isolation system cabinet caused three main steam isolation valves (MSIVs) to shut resulting in a reactor trip from approximately 100 percent power. The MSIVs closed as a result of a fire in the main steam and feedwater isolation system cabinet. The fire damaged a logic module that contained both normal control and safety-related protection features. During the event, MSIV-180 inadvertently reopened and the operators were unable to remotely close the valve. The valve was later closed by local-manual operation.
Condition Reports/Disposition Request 2405660 documented this event. The associated equipment root cause failure analysis report, approved on January 17, 2002, concluded that the logic design allowed the fire damage to the logic module to re-open MSIV-180. In addition, the licensee determined that MSIV-180 may not have closed on a main steam isolation actuation signal. The licensees report stated that the logic design did not meet the requirements of IEEE-279, Section 4.2, Single Failure Criterion, Section 4.6, Completion of a Protective Action Once it is Initiated, and Section 4.7, Control and Protection System Interaction. The Final Safety Analysis Report commits the licensee to IEEE-279.
The licensee completed an immediate operability evaluation and on August 31, 2001, put temporary compensatory measures in place. A subsequent formal operability evaluation (Operability Determination 246) was completed January 29, 2002, and additional compensatory measures were instituted.
The team was concerned about the length of time between the event occurrence and completion of the root cause failure analysis report. This delay was potentially due to contradicting engineering views on the conformance of the main steam and feedwater isolation system to design requirements. The initial equipment root cause failure analysis report submitted for approval on August 17, 2001, determined that the logic design for the MSIV control circuitry did not comply with the requirements of IEEE-279.
Approval of this report was delayed until January 13, 2002. On October 9, 2001, as documented in Condition Reports/Disposition 2418186, the licensee determined that this system conformed to design requirements. On January 25, 2002, the disparity in the two evaluations was resolved when the engineering director concurred with the position that the main steam and feedwater isolation system did not meet IEEE-279 requirements. The team noted that the formal operability determination was not completed until 6 months after the event.
The team concluded that sufficient compensatory actions were taken shortly after the event to ensure that safety functions of the main steam and feedwater isolation system would not be defeated following a fire. However, this licensee event report remains open pending NRC review of design controls associated with IEEE-279 requirements and potential modifications to the MSIV control circuitry.
(Closed) Licensee Event Report 50-530/2001-003-00: Leak in an Inconel Alloy 600 Instrument Nozzle in the Reactor Coolant System. The team reviewed Condition Reports/Disposition Request 2427919. Evidence of a leak was found on Reactor Coolant System Loop 1B hot-leg instrument nozzle during a routine inspection. The team reviewed Condition Reports/Disposition Request 2427919. The nozzle was repaired using a mechanical nozzle seal assembly. The leak was attributed to primary water stress corrosion cracking. The leakage was on the order of ounces per year and was anticipated. The licensees analysis indicated that this type of leak will become evident through small cracks prior to significant degradation of the pressure boundary.
The licensee intends to replace this nozzle and remaining Alloy 600 hot-leg instrument nozzles in all three units with nozzles fabricated from an alloy less susceptible to this type of degradation during Outages R8 to R10 for each unit, respectively. No findings of significance were identified. This licensee event report is closed.
(Closed) Licensee Event Report 50-528, 529, 530/2001-S001-00: Licensee Denied Access to an Individual who had Previously been Granted Unescorted Access Based on Pre-Employee Screening Records. The team reviewed Condition Reports/Disposition Request 2373714 and interviewed the Emergency Services Department Programs Department Leader. It was discovered that an employee had failed to disclose an arrest for possession of a controlled substance. This was discovered during review of the individuals FBI records and a subsequent investigation, which included an interview with the contractor employee when he returned after a several month absence in February 2001. Access had previously been terminated on October 30, 2000, after making eight entries in one month. Because the arrest was still active, the individual was denied access. The licensee checked his work assignments and found that he had performed no safety-related work. The licensee added additional information to PADS in the event the individual attempted to enter another licensed facility. No findings of significance were identified. This licensee event report is closed.
(Open) Licensee Event Report 50-528, 529, 530/2001-003-00: Technical Specifications Required Shutdown Due to Degraded Control Element Assemblies (CEAs). The degradation involved cracks in the fingers containing the boron carbide poison material and loss of boron carbide pellets into the reactor coolant system in some instances.
The team reviewed Condition Reports/Disposition Requests 2427919, 2377444, 2412913, 2375404, and 2376822. The team also interviewed engineers and managers engaged in addressing the issues associated with the degraded CEAs. The team determined that the licensee understood the failure mechanism, the location of degraded material, and the reactivity effects of the CEA degradation. There was no safety system functional failure. The reactors maintained the shutdown reactivity assumed in the safety analysis report at all times. All loose material resulting from the failures had been appropriately recovered or accounted for.
As a part of the corrective action, the licensee had replaced all CEAs in all three units.
They had inspected all the removed CEAs and had identified some cracking in the fingers of all assemblies. They had determined that the expected lifetime for CEAs was much shorter than the 12 cycles estimated by the vendor - probably on the order of 6 cycles - and had engaged the vendor in attempting to determine a new expected
lifetime. Because a foreign reactor was going into a refueling outage after 6 cycles of operation with similar CEAs, the licensee was working with the vendor to thoroughly inspect these CEAs as a part of the effort to establish a new CEA expected lifetime.
Similar CEAs are not used in other domestic reactors. No findings of significance were identified. This licensee event report remains opening pending determination of a new expected lifetime for CEAs.
4OA6 Meetings, including Exit Exit Meeting An exit meeting was held on February 1, 2002, with W. Ide, Vice President, Nuclear Production, and other licensee staff members during which the team leader characterized the results of the inspection. The licensee's management acknowledged the findings presented. On March 19, 2002, Mr. Anthony T. Gody informed M. Sontag during a telephone conversation, that Licensee Event Report 2-2001-002-00 would remain open.
The team leader asked the licensee's management whether any materials examined during the inspection should be considered proprietary. No proprietary information was identified.
ATTACHMENT KEY POINTS OF CONTACT Licensee K. Atkinson, Operations Shift Manager B. Bandera, Department Leader, Nuclear Fuel Analysis S. Burns, System Engineering Department Leader T. Cahill, Senior Engineer, Nuclear Analysis D. Douglass, Senior Evaluator, Nuclear Assurance Department E. Dutton, Nuclear Assurance Department, Section Leader A. David Huttie, Programs Department Leader, Emergency Services Department M. Hypse, System Engineering Section Leader W. Ide, Vice President, Nuclear Production D. Kanitz, Nuclear Regulatory Affairs, Senior Engineer J. Levine, Executive Vice President, Generation B. Lindenlaub, Engineer D. Marks, Regulatory Affairs-Compliance Section Leader G. Overbeck, Senior Vice President, Nuclear N. Pappas, Operations Shift Manager J. Roland, Senior Engineer., Nuclear Assurance Department R. Quesinberry, Maintenance Senior Advisor C. Seaman, Nuclear Regulatory Affairs, Director E. Sonn, Significant Investigator M. Sontag, Corrective Action Program, Department Leader N. Thibodaux, Senior Consulting Engineer D. Vogt, STA Section Leader NRC J. Moorman, Senior Resident Inspector ITEMS CLOSED AND DISCUSSED Closed 3-2001-003-00 LER Leak in an Inconel Alloy 600 Instrument Nozzle in the Reactor Coolant System (Section 4OA3)
1,2,3-2001-S001-00 LER Licensee Denied Access to an Individual who had Previously been Granted Unescorted Access Based on Pre-Employee Screening Records (Section 4OA3)
Discussed 1, 2, 3-2001-003-00 LER Technical Specifications Required Shutdown Due to Degraded Control Element Assemblies (Section 4OA3)
2-2001-002-00 LER Logic Board and Pin Connector Failure Causes Three of Four Main Steam Isolation Valves to Close (Section 4OA3)
DOCUMENTS REVIEWED The following documents were selected and reviewed by the inspectors to accomplish the objectives and scope of the inspection and to support any findings:
PROCEDURES 40DP-9OP26, Operability Determination, Revision 10 30DP-9WP02, Work Document Development and Control, Revision 29 90DP-0IP10, Condition Reporting, Revision 13 30DP-0RA01, Component Failure Trending, Revision 4 81DP-0DC13, Deficiency (DF) Work Order, Revision 13 60DP-0QQ02, Trend Analysis and Coding, Revision 11 12DP-0MC29, Warehouse Discrepancy Notice (WD), Revision 11 70DP-0EE01, Equipment Root Cause of Failure Analysis, Revision 10 40OP-9SG01, Main Steam, Revision 23 CRDR Processing Guideline, Revision 3 40DP-9OP02, Rev. 17, Conduct of Shift Operations 40DP-9OP22, Rev. 15, Operations Log Keeping 72OP-9RX01, Rev. 7, Calculation of Estimated Critical Condition 72OP-9RX02, Rev. 2, Determination of Anticipated Critical Position 40OP-9ZZ03, Rev. 20, Reactor Startup CONDITION REPORT AND DISPOSITION REQUESTS (CRDRS)
0117562 0239183 117555 2305275 2331089 2332280 2333473 2339523 2347599 2350293 2352119 2360498 2365447 2368394 2369601 2369999 2370347 2372245 2373569 2373714 2374606 2374701 2375404 2376079 2376822 2377444 2383862 2384347 2384489 2385849 2388632 2391526 2392546 2393348 2394824 2400881 2405660 2406836 2412307 2412310 2412913 2413687 2414063 2414747 2414776 2414777 2414867 2417046 2418186 2419385 2421912 2424725 2425046 2425046 2425946 2426366 2427901 2427901 2427919 2430432 2430974 2432005 2432646 2436184 2436262 2436446 2436934 2437828 2442947 2444131
OPERABILITY DETERMINATIONS 2356074 2362235 2373313 2401544 2407828 2417365 2417545 WORK ORDERS 211137 211142 218988 211071 237640 216417 1107803 951668 227740 1080587 219118 217121 216241 219358 219359 211191 218745 218740 219401 LICENSEE EVENT REPORTS 1-2001-001, Boric Acid on Unit 1 Reactor Coolant System Hot Leg Instrument Nozzle, May 24, 2001 2-2000-004, Reactor Coolant System Pressure Boundary Leakage Due to Degraded Alloy 600 Pressurizer Heater Sleeve, November 1, 2000 1-2001-002, LLRT Methodology may Not have Correctively Quantified Leakage for Inboard Containment Isolation Valves, May 24, 2001 1-2000-004, Technical Specification Violation Due to Deficient Test Procedure for Refueling Purge Valves, December 28, 2000 1, 2, 3-2001-003, Cracks found in Unit 3 CEA, December 4, 2001 1-2001-004, Technical Specifications Violation for Inoperable RCS Leak Detection because of Misaligned O-Ring, December 12, 2001 1-2001-S01, Licensee Denied Access to an Individual who had Previously been Granted Unescorted Access Based on Pre-Employee Screening Records, April 18, 2001 2-2000-008, Lift Values on Two of Four Pressurizer Safety Valves were Outside of Technical Specifications Limits, January 10, 2001 2-2000-009, Main Steam Safety Valve Lift Pressure outside of Technical Specifications Limits, February 16, 2001 2-2001-001, Main Steam Safety Valve Lift Pressure outside of Technical Specifications Limits, May 24, 2001 2-2001-002, Logic Board and Pin Connector Failure Causes Three of Four Main Steam Isolation Valves to Close, August 31, 2001
3-01-001, Reactor Tripped from 19 Percent Rated Thermal Power from Axial Shape Index Auxiliary Trip Signal from Core Protection Calculator, July 18, 2001 3-01-003, Leak in an Inconel Alloy 600 Instrument Nozzle in the Reactor Coolant System, November 28, 2001 MISCELLANEOUS Palo Verde System Health Report - 3rd Quarter 2001" Equipment Root Cause Failure Analysis Program Assessment, August 17, 2000 Night Order, MSFIS and FWIV Operation, February 1, 2002 Operability Determination #246, MSIV and FWIV Operation, January 31, 2002 Audit Report 2001-008, Corrective Action NONCITED VIOLATIONS NCV 01-03-03, AFW Pump Made Inoperable when Steam Trap Removed from Service and NOT Operated per Procedure
Material Request for Palo Verde Problem Identification and Resolution Inspection 1.
A summary list of all currently open/active items for:
CRDRs of significant conditions adverse to quality operator work-arounds engineering review requests temporary modifications procedure change requests training needs request/evaluation control room and safety system deficiencies human performance issues 2.
A summary list of all items completed/resolved/closed since January 1, 2001 for:
CRDRs of significant conditions adverse to quality operator work-arounds engineering review requests temporary modifications procedure change requests training needs request/evaluation control room and safety system deficiencies human performance issues 3.
Summary list of all CRDRs generated during the specified period and sorted by:
chronology initiating organization responsible organization 4.
All quality assurance audits and surveillance of corrective action activities since January 1, 2001.
5.
All corrective action activity resulting from functional area self-assessments and Non-NRC third party assessments since January 1, 2001.
6.
Corrective action performance trending/tracking reports generated since January 1, 2001.
g.
Current revision of the procedures governing initiation and processing of CRDRs, potential conditions adverse to quality, and root cause analysis.
8.
Any additional governing procedures/policies/guidelines for:
Condition Reporting Corrective Action Program Root Cause Evaluation/Determination
Operator Work-Arounds Work Requests Engineering Requests Temporary Modifications Procedure Change Requests Deficiency Reporting and Resolution Training Needs Request/Evaluation h.
For each of the items applicable to Palo Verde listed below please provide the following:
Full text of the CRDRs (please indicate any findings that did not result in a CRDR or corrective actions)
Any "Roll-up" or "Aggregating" CRDRs related to the generic communication or condition report.
- Root Cause analysis report (if applicable)
Risk significance assessments
Probable Cause evaluation (if applicable)
Approved corrective actions
Basis for extending originally approved due dates
Evidence of corrective action completion for those items deemed to be closed (work packages, design change documentation, temporary modifications, training lesson plans/material, training attendance records, procedure revisions, etc.)
10.
Part 21 Reports:
2001-01-0: 12/12/00 710DUOCL calibration unit
2001-02-0: 12/15/00 Unrecognized capacitor orientation
2001-03-0: 12/18/00 Seismic qualification of electrically operated AK-15/25 circuit breakers
2001-04-0: 01/04/01 weights found in removed SFP storage racks
2001-05-0: 12/07/00 Defective weld in HL near vessel
2001-06-0: 12/20/00 GTSTRUDL dynamic analysis command
2001-07-0: 01/10/01 Potential EDG inoperability for Agastat relays
2001-08-0: 01/15/01 medium voltage circuit breaker failures
2001-09-0: 01/18/01 segregation of ingredients in safety-related grout
2001-10-0: 01/31/01 broken cap screw in aux. feedwater pump
2001-11-0: 02/28/01 internal binding of terminal shaft in Woodward governors
2001-12-0: 02/28/01 relay label mismatch LER 50-293/2001-01
2001-13-0: 03/28/01 breaker cubicle mechanism out of spec.
- 2001-14-0: 03/29/01 inappropriate reference temperature used
2001-15-0: 04/16/01 replacement Foxboro differential pressure
2001-16-0: 04/10/01 broken bases in CV-7 relays
2001-17-0: 04/27/01 calculation of time to criticality
2001-18-0: 05/02/01 Failure of K-Line circuit breaker to close
2001-19-0: 05/11/01 R-11 radiation monitor spiking
2001-20-0: 05/23/01 low flow coefficients for ball check valves
2001-21-0: 06/19/01 electrolytic capacitors in Woodward 2301A control devices
2001-22-0: 06/21/01 leaking flow switch in containment gas analyzer
- 2001-24-0: 07/09/01 Woodward governor exhibits unstable oscillations
2001-25-0: 08/08/01 Nine Mile Point Unit 1 incompletely threaded screw on terminal block
2001-26-0: 08/13/01 Scientech temperature difference module yields current instead of voltage output Event Notification 38204
- 2001-26-1: 09/12/01 Scientech temperature difference module yields current instead of voltage output
- 2001-27-0: 08/23/01 Flowserve excessive disc angular movement in swing check valves
- 2001-28-0: 08/24/01 Rosemount non-qualified screws used for remote seals in pressure transmitter
- 2001-29-0: 08/31/01 David Brown Union Pumps potential loss of backup safety function of charging pump air lock tank 11. NRC Information Notices:
2001-001: 03/26/01 Importance of accurate inventory controls to prevent the unauthorized possession of radioactive material
2001-002: 03/28/01 Summary of fitness-for-duty program performance reports for calendar years 1998 and 1999
2001-004: 04/11/01 Neglected fire extinguisher maintenance causes fatality
2001-005: 04/30/01 Thru-wall cracking of RPV head control rod drive mechanism penetration nozzles
2001-006: 05/11/01 Centrifugal charging pump thrust bearing damage not detected
2001-007:
05/11/01 Unescorted access granted on the basis of incomplete/inaccurate information
2001-009: 06/12/01 Main FW system degradation in safety-related ASME code class 2 piping inside containment of a PWR
2001-010: 06/28/01 Failure of Central Sprinkler Co. Model GB Series fire sprinkler heads
2001-012: 07/13/01 Hydrogen fire at a Nuclear Power Station
2001-13: 08/10/2001 NRC Information Notice Inadequate Standby Liquid Control System Relief Valve Margin 12.
All NCVs and NOVs issued since January 1, 2001 13.
Current System Health Reports or similar system information 14.
Listing of plant safety issues generated through the employee concerns program since January 1, 2001 15.
Listing of action items generated by the plant safety review committees since January 1, 2001 16.
Current predictive performance summary reports 17.
All LERs generated for the three units since January 1, 2001.