IR 05000219/2004006
| ML041830367 | |
| Person / Time | |
|---|---|
| Site: | Oyster Creek |
| Issue date: | 07/01/2004 |
| From: | Ray Lorson NRC/RGN-I/DRS/PEB |
| To: | Crane C AmerGen Energy Co |
| Lorson R, RI/DRS/PEB, (610) 337-5282 | |
| References | |
| -RFPFR IR-04-006 | |
| Download: ML041830367 (22) | |
Text
July 1, 2004
SUBJECT:
OYSTER CREEK GENERATING STATION - NRC PROBLEM IDENTIFICATION AND RESOLUTION INSPECTION REPORT 05000219/2004006
Dear Mr. Crane:
On May 21, 2004, the US Nuclear Regulatory Commission (NRC) completed a team inspection at the Oyster Creek Generating Station. The enclosed report documents the inspection findings that were discussed on May 21, 2004, with Mr. C. N. Swenson and other members of your staff during an exit meeting.
This inspection was an examination of activities conducted under your license as they relate to the identification and resolution of problems, and compliance with the Commissions rules and regulations and the conditions of your operating license. Within these areas, the inspection involved examination of selected procedures and representative records, observation of activities, and interviews with personnel.
On the basis of the samples selected for review, the team concluded that in general, problems were properly identified, evaluated, and corrected. The team identified one finding of very low safety significance (Green) associated with the corrective actions for a deficiency associated with operation of the reactor mode switch during a reactor trip on August 14, 2003. The finding was determined to be a violation of NRC requirements. However, because of its very low safety significance and because it was entered into your corrective action program, the NRC is treating this finding as a non-cited violation (NCV) consistent with Section VI.A of the NRC Enforcement Policy. If you deny this NCV, you should provide a response with the basis for your denial within 30 days of the date of this inspection report, to the U. S. Nuclear Regulator Commission, ATTN. Document Control Desk, Washington, D.C. 20555-0001, with copies to the Regional Administrator, Region I; the Director, Office of Enforcement, U. S. Nuclear Regulator Commission, Washington, D.C. 20555-0001; and the NRC Resident Inspector at the Oyster Creek Generating Station. In accordance with 10 CFR 2.390 of the NRC's "Rules of Practice,"
a
Mr. Christopher 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/
Raymond K. Lorson, Chief Performance Evaluation Branch Division of Reactor Safety Docket No: 50-219 License No: DPR-16 Enclosure:
Inspection Report 05000219/2004006 w/Attachment: Supplemental Information cc w/encl:
Chief Operating Officer, AmerGen Site Vice President, Oyster Creek Nuclear Generating Station, AmerGen Plant Manager, Oyster Creek Generating Station, AmerGen Regulatory Assurance Manager Oyster Creek, AmerGen Senior Vice President - Nuclear Services, AmerGen Vice President - Mid-Atlantic Operations, AmerGen Vice President - Operations Support, AmerGen Vice President - Licensing and Regulatory Affairs, AmerGen Director Licensing, AmerGen Manager Licensing - Oyster Creek, AmerGen Vice President, General Counsel and Secretary, AmerGen T. ONeill, Associate General Counsel, Exelon Generation Company J. Fewell, Assistant General Counsel, Exelon Nuclear Correspondence Control Desk, AmerGen J. Matthews, Esquire, Morgan, Lewis & Bockius LLP Mayor of Lacey Township K. Tosch - Chief, Bureau of Nuclear Engineering, NJ Dept. of Environmental Protection R. Shadis, New England Coalition Staff N. Cohen, Coordinator - Unplug Salem Campaign W. Costanzo, Technical Advisor - Jersey Shore Nuclear Watch E. Zobian, Coordinator - Jersey Shore Anti Nuclear Alliance
Mr. Christopher
SUMMARY OF FINDINGS
IR 05000219/2004006; 04/26 - 04/30/04 and 05/17 - 05/21/04; Oyster Creek Generating
Station; biennial baseline inspection of the identification and resolution of problems. One violation was identified in the area of corrective actions.
This inspection was conducted by two regional inspectors and three resident inspectors. The inspection identified one Green finding that was a non-cited violation of NRC requirements.
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.
Identification and Resolution of Problems Based on the sample items selected for review, the team concluded the implementation of the corrective action program at Oyster Creek Generating Station was adequate. The team determined that AmerGen was generally effective at identifying discrepant conditions at an appropriate threshold and entering them into the corrective action program. Identified issues were typically prioritized appropriately and in a timely fashion and were properly evaluated commensurate with the potential safety significance. Overall, the evaluations reasonably identified the causes of the problem, the extent of the condition, and provided for corrective actions to address the causes. However, in some cases, the corrective action program was not effectively used to evaluate, resolve and prevent problems. There were also some examples where issue evaluations were not complete, and corrective actions were not effective at resolving problems. Audits and self-assessments identified adverse conditions and negative trends, and were generally self-critical and consistent with the teams findings. On the basis of interviews conducted, the team determined that plant staff personnel were familiar with and utilized the corrective action program to identify problems.
NRC Identified and Self-Revealing Findings
Cornerstone: Mitigating Systems
- Green.
The team identified a non-cited violation of 10 CFR 50, Appendix B,
Criterion XVI, Corrective Actions, which requires that prompt corrective actions be implemented for conditions adverse to quality. Specifically, AmerGen did not implement a planned corrective action to address a deficiency associated with operation of the reactor mode switch during a reactor trip on August 14, 2003.
The finding was determined to be more than minor because it negatively affected the mitigating systems cornerstone attribute of human performance. Failure to place the reactor mode switch into the shutdown position following a reactor scram would be expected to result in a loss of the normal heat sink and complicate the event response. The finding was of very low safety significance (Green), because it was not a design or qualification deficiency, and it did not iii result in an actual loss of safety function for risk-significant equipment with respect to internal or external events. Additionally, the team noted that the heat sink would be recoverable from an event of this type. (Section 4OA2.c.2.1)
REPORT DETAILS
OTHER ACTIVITIES (OA)
4OA2 Problem Identification and Resolution
a.
Effectiveness of Problem Identification
- (1) Inspection Scope The team reviewed AmerGens corrective action program and noted that problems were formally identified through the initiation of action requests (ARs) or corrective action program reports (CAPs). To understand the threshold for identifying problems and to assess management involvement with the corrective action process, team members attended daily work management meetings where ARs were reviewed for disposition and assignment, and daily screening and management review committee meetings where CAPs were screened for significance and assignment. The team also selected items from AmerGens nuclear oversight (NOS) and focused area self-assessment (FASA) processes to verify that AmerGen appropriately considered problems identified through these processes for entry into the corrective action program. Specifically, the team reviewed a sample of control room deficiency and work-around lists, operability evaluations, system health reports, maintenance orders, and NOS audits and FASA reports.
The team reviewed selected ARs and CAPs initiated subsequent to the last problem identification and resolution (PI&R) inspection completed in June 2002 to determine whether AmerGen was appropriately identifying, characterizing, and entering problems into the corrective action process. The team selected ARs and CAPs to cover the seven cornerstones of safety identified in the NRC reactor oversight process (ROP). The team used the individual plant examination (IPE) report, site-specific SDP worksheets, and individual system performance indicators to focus system walkdowns and AR and CAP sample selection. The team focused its review of AmerGens corrective actions on the following systems: emergency diesel generators (EDGs), containment spray/emergency service water (CS/ESW), reactor building component cooling water (RBCCW),instrument air (IA), and 1E 125 Vdc. The attachment lists the ARs and CAPs selected for review.
The team interviewed selected plant staff to determine whether personnel were familiar with and utilized the corrective action program to identify problems. The team also conducted walkdowns of the control room panels and the selected systems to verify that problems were identified and addressed at an appropriate level.
- (2) Observations and Findings No findings of significance were identified.
The team determined that, in general, AmerGen adequately identified discrepant conditions and initiated CAPS or ARs where appropriate. Audits and self-assessments identified adverse conditions and negative trends, and were generally self-critical and consistent with the teams findings. However, the team noted several examples where AmerGen did not enter conditions adverse to quality into the corrective action system and/or did not identify and correct other minor deficiencies in a timely manner. In response to observations made during the teams plant walkdowns and documentation reviews, AmerGen generated 15 CAPs. Some of these issues included:
- The team identified that the simulator key-lock reactor mode switch allowed switch operation with the switch locked and the key removed. This was not consistent with the operation of the mode switch in the control room and could have provided negative training regarding reactor mode switch operations.
- An IA sample result indicated that the total hydrocarbon content was above the air quality specification used by AmerGen, however, no CAP was written to evaluate the issue.
- AmerGen previously issued CAP O2004-0909 to evaluate and correct excessive IA leakage. During a system walkdown the team identified two previously unidentified air leaks from a pipe union and pipe cap. AmerGen investigated the leaks, but failed to analyze the impact of the additional leaks on the operation of the system.
- RBCCW return flow from the drywell was required to be throttled to maintain RBCCW pressure in the drywell piping. The system operating procedure directed the use of a gate valve to throttle the return flow from the drywell. In interviews with the team, operators stated that flow control using this method was difficult and flow could take as long as 15 minutes to stabilize following valve adjustments; however, no CAP or AR was written to evaluate the issue.
- AmerGen did not initiate a CAP for an issue identified during a December 2001 FASA involving a fire protection drawing error in the 480 Vac switchgear room, and also did not initiate CAPs for deficient control room indications that were identified during a March 2004 corrective action program FASA.
AmerGen subsequently issued CAPs or ARs for each of the identified issues. The team independently evaluated the problem identification deficiencies noted above for potential significance. The team determined that none of the individual issues were findings of more than minor significance based upon the guidance in Inspection Manual Chapter (IMC) 0612, Appendix E, Examples of Minor Issues.
b.
Prioritization and Evaluation of Issues
- (1) Inspection Scope The team reviewed the ARs and CAPs listed in the attachment to determine whether AmerGen adequately evaluated and prioritized problems. The review included the appropriateness of the assigned significance, the timeliness of resolutions, and the scope and depth of the root cause analyses. The ARs and CAPs reviewed encompassed the full range of AmerGen evaluations, including root and apparent cause evaluations. The team selected the ARs and CAPs to cover the seven cornerstones of safety identified in the NRC ROP. A portion of the items chosen for review were those that were age dependent, and accordingly, the scope of review was expanded to five years. In this area, the team reviewed items associated with service water system piping degradation and cracking of vertical welds in the core shroud. The team used risk insights from AmerGens IPE to focus the AR and CAP sample to the RBCCW, CS/ESW, EDGs, IA and 1E 125 Vdc systems. Additionally, the team attended the daily management meeting to observe the review process and to understand the basis for assigned significance levels.
The team also selected a sample of CAPs associated with previous NRC findings to determine whether AmerGen evaluated and resolved problems associated with compliance to applicable regulatory requirements and standards. The team reviewed AmerGens assessment of equipment operability, reportability requirements, and extent of condition. The team reviewed AmerGens evaluation of industry operating experience (OE) information for applicability to their facility. The team also reviewed AmerGens response to NRC identified issues during the inspection.
- (2) Observations and Findings No findings of significance were identified.
The team determined that, in general, AmerGen adequately prioritized and evaluated the issues and concerns entered into the corrective action program. Personnel were generally effective at classifying and performing operability evaluations and reportability determinations for discrepant conditions. However, the team identified weaknesses with some of AmerGens evaluations of conditions adverse to quality. Some examples included:
- CAP O2003-1681 documented that during surveillance testing one main steam isolation valve (MSIV) stroked faster than the surveillance test acceptance criterion. The team determined that during the testing, operators stroked the valve four times by repeating the procedure, and that three of the four times the valve stroked faster than the test acceptance criterion. AmerGen did not evaluate whether the repetitive fast stroking had any adverse impact on the valve. AmerGen issued CAP O2004-1230 for this NRC identified issue.
- During a tour of the 480 Vac switchgear room the team identified a loose floor plug. The fire protection engineers evaluated the condition of the plug and determined that the issue did not require entry into the corrective action program.
However, the engineers did not formally evaluate the impact of the loose plug on the rooms halon suppression system until questioned by the team. AmerGen issued CAP O2004-1205 for this NRC identified issue.
The team noted that there were no significant adverse consequences or operability issues associated with these observations, and AmerGen initiated CAPs to address both conditions. The team also reviewed IMC 0612, Appendix E, Examples of Minor Issues, and determined that these corrective action performance deficiencies were of minor significance and not subject to formal enforcement action.
c.
Effectiveness of Corrective Actions
- (1) Inspection Scope The team reviewed AmerGens corrective actions for the selected ARs and CAPs listed in the report attachment to determine whether actions addressed the identified causes.
The team reviewed AmerGens timeliness in implementing corrective actions and their effectiveness in preventing recurrence of significant conditions adverse to quality. This review included CAPs associated with the NCVs and findings issued since the last PI&R inspection, to determine whether AmerGen properly evaluated and resolved these issues.
The team reviewed a sample of control room deficiency and work-around lists, operability evaluations, system health reports, maintenance orders, and NOS audits and FASA reports to confirm that the underlying problems associated with each issue were properly resolved.
In addition the team assessed AmerGens backlog of corrective actions to determine, if any, individually or collectively, represented an increased plant risk due to the delay in implementation.
- (2) Observations and Findings One Green finding was identified involving the failure to implement a planned corrective action for a deficiency associated with operation of the reactor mode switch during a reactor trip on August 14, 2003. In addition, the team noted some examples where AmerGens resolution of degraded conditions, documentation of actions, and completion of identified corrective actions were not fully effective. Specifically:
- The station was slow to address NOS identified issues associated with implementation of the corrective action program as discussed in CAP 2004-0148. The issues were identified in early 2003, but were not resolved prior to the end of 2003. In February 2004, NOS issued an elevation notice that required a formal response from the corrective action group for these issues.
- The team identified several weaknesses associated with the corrective actions for CAP O2002-1551 which documented a void condition below the 480 Vac switchgear rooms. This condition affected the separation of redundant 4160 Vac safety-related cable trains. Amergens corrective actions included drilling holes in the 480 Vac switchgear room floors and using these holes to backfill the void area with sand. The team noted that the 480 Vac switchgear room floor had been classified as a three hour barrier in the fire hazards analysis, but Amergen did not complete a formal evaluation prior to drilling these holes and did not produce any formal data to demonstrate that the as left condition was equivalent to a three hour barrier. Additionally, AmerGen performed an analysis per Generic Letter (GL) 86-10 to demonstrate that the as left condition met applicable Appendix R requirements. However, the analysis did not provide a rigorous technical basis that a postulated event would not affect multiple safe-shutdown trains.
- Another observation by the team related to AmerGens evaluation and corrective actions for NCV 2002008-003, which dealt with an inadequate in-service test procedure for the 52B ESW pump. The team determined that the evaluation and corrective actions did not fully address the issue documented in the inspection report. As a result, the scope of the issue was not fully identified and accordingly, the appropriate corrective actions were not taken. The team noted no significant adverse consequences or operability issues associated with this observation, and AmerGen initiated CAP O2004-1111 to address the problem.
The team independently evaluated the corrective action program deficiencies noted above for potential significance. The team determined that none of the individual issues were findings of more than minor significance based upon the guidance in IMC 0612, Appendix E, Examples of Minor Issues. However, these issues represented examples where the corrective actions for identified conditions were not fully effective.
.1 Failure to Implement Adequate Corrective Actions for a Reactor Mode Switch
Operational Problem
Introduction.
A Green NCV was identified for failure to implement adequate corrective actions for a reactor mode switch operational problem, as prescribed in 10 CFR 50, Appendix B, Criterion XVI.
Description.
Following a reactor trip on August 14, 2003, two control room operators experienced difficulty in moving the reactor mode switch from the run position to the shutdown position. This allowed an automatic closure of the main steam isolation valves (MSIVs) to occur when the main steam line pressure fell below the low pressure setpoint. Closure of the MSIVs represented a loss of the normal heat sink and complicated the response to this event.
AmerGens post-transient event review identified that the reactor mode switch locking mechanism had a history of "binding" when operated (CAP O2003-1621). Their causal analysis attributed the operators inability to re-position the mode switch in a timely manner to either a mechanical problem or operator error. The planned corrective actions for this event included: maintenance on the switch, development of a shift training brief, and enhanced training on mode switch operations in the licensed operator training programs. AmerGen completed the corrective actions associated with the maintenance checks on the switch and with the shift training brief, but did not implement the corrective action associated with enhancement of the licensed operator training programs.
The team noted that no mechanical problems were identified with the switch during the maintenance activities and concluded that operator error was the most likely cause for the delayed operation of the mode switch during the August 14, 2003 event. The team also interviewed operators and operator training personnel and learned that the switch was sometimes tricky or difficult to operate. The team concluded, based on the above, that the corrective actions taken for CAP O2003-1621 were not sufficiently thorough to ensure that operators could successfully operate the reactor mode switch during an event.
Analysis.
Following an August 14, 2003 reactor trip, AmerGen did not implement adequate corrective actions to ensure that operators would be able to successfully operate the reactor mode switch during an event. This is a performance deficiency.
Traditional enforcement does not apply because the issue did not have any actual safety consequences or potential for impacting the NRCs regulatory function and was not the result of any willful violation of NRC requirements or AmerGen procedures.
The finding was more than minor because it adversely affected the mitigating systems cornerstone attribute of human performance. Specifically, difficulty in placing the reactor mode switch to the shutdown position in a timely manner following a reactor scram could lead to a loss of the normal heat sink and complicate the event response.
Therefore, this deficiency affected the availability of a system that responds to initiating events to prevent undesirable consequences. In accordance with IMC 0609, Appendix A, Significance Determination of Reactor Inspection Findings for At-Power Situations, the team determined that the finding was of very low safety significance, because it was not a design or qualification deficiency, and it did not result in an actual loss of safety function for risk-significant equipment with respect to internal or external events.
Additionally, the team noted that the normal heat sink was recoverable from this type of event. AmerGen entered this finding into their corrective action program as CAP O2004-1253.
Enforcement.
10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, requires that measures shall be established to assure that conditions adverse to quality, such as failures, malfunctions, deficiencies, deviations, defective material and equipment, and non-conformances are promptly identified and corrected. Contrary to the above, following the August 14, 2003 reactor trip, AmerGen did not implement adequate corrective actions to ensure proper operation of the reactor mode switch. Because the failure to correct this condition adverse to quality is of very low significance and has been entered into AmerGens corrective action program (CAP O2004-1253), this violation is being treated as an NCV, consistent with Section VI.A of the NRC Enforcement Policy, issued May 1, 2000 (65FR25368). (NCV 05000219/2004003-01)
4OA6 Meetings, including Exit
The team presented the inspection results to Mr. C. Swenson and other members of AmerGen management on May 21, 2004. AmerGen management acknowledged that no proprietary information was involved.
ATTACHMENT
SUPPLEMENTAL INFORMATION
Licensee Personnel
- J. Hackenberg, Manager - Operations Training
- M. Godknecht, Maintenance Rule Coordinator
- J. Magee, Director, Engineering
- M. Massaro, Plant Manager
- D. McMillan, Director, Training
- L. Newton, Manager, Chemistry & Rad Protection
- J. Renda, Manager - Radiation Protection
- D. Slear, Manager, Regulatory Assurance
- B. Stewart, Senior Licensing Engineer
- C. Swenson, Vice President
- C. Wilson, Director, Operations
- P. Cervanka, Manager, Nuclear Oversight
- M. Taylor, Employee Concerns Program
- J. Freeman, Shift Operations Superintendent
- D. Chernesky, Manager Mechanical/Electrical Maintenance
LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED
Opened and Closed
Following the August 14 reactor scram AmerGen failed to
take adequate corrective action to ensure operators can
successfully operate the reactor mode switch during an
event. (Section 4OA2.c.2.1)
LIST OF DOCUMENTS REVIEWED
Procedures:
101.2
Fire Protection Program, Revision 51
331.1
Control Room and Old Cable Spreading Room Heating, Ventilation, and
Air Conditioning System, Revision 15
607.4.016
Containment Spray and Emergency Service Water System 1 Pump
Operability and Quarterly Inservice Test, Revision 0
DD-11
Nuclear Oversight Department Description, Revision 5
Operability Determinations, Revision 1
Detailed Operability Determination, Revision 0
Root Cause Analysis Manual, Revision 4
Common Cause Analysis Manual, Revision 3
Apparent Cause Evaluation Manual, Revision 4
Effectiveness Review Manual, Revision 2
CAP Process Expectations Manual, Revision 5
LS-OC-125
Corrective Action Program (CAP) Procedure, Revision 3
Independent Safety Engineering Function Process Description, Revision
Nuclear Oversight Training, Qualification, and Certification, Revision 5
OP-OC-100
Oyster Creek Conduct Of Operations, Revision 1
CC-OC-112-1001
Temporary Configuration Change Implementation, Revision 2
2400-GME-3780.52
Installation, Testing And Termination Of Wire And Cable, Revision 7
Dosimetry Issue, Usage, and Control, Revision 4
RP-OC-2001
Dosimetry Investigation Reports, Revision 0
607.4.015
Containment Spray & ESW System-2 Pump Operability Quarterly IST,
Revision 0
607.4.016
Containment Spray & ESW System-1 Pump Operability Quarterly IST,
Revision 0
607.4.004
Containment Spray & ESW System-1 Pump Operability Comprehensive,
Pre-service, Post Maintenance In-service Test, Revision 52
2.10
Secondary Containment Control, Revision 8
636.4.016
Diesel Generator No. 2 Fast Start Test, Revision 1
Non-Cited Violations Reviewed:
2003-03-01
Failure to Adequately Maintain the ESW System as Required by technical specification 6.8.1
2002-07-03
Ineffective Corrective Actions for Preventing Over-Pressure in control rod
drive system
2003-05-01
Failure to implement a surveillance test procedure required by technical specifications 6.8.1.B
2003-03-05
License violation due to security officer inattentive to duty
2001-13-02
Failure to take corrective actions to preclude repetition of a significant
condition adverse to quality associated with safety related 480 volt
electrical circuit breakers.
2002-03-01
Failure to promptly identify and correct a low air flow condition in A control
room ventilation
2002-08-02
Failure to identify and correct a degraded condition with respect to the
standby gas treatment system charcoal filters
2002-08-01
Failure to properly implement engineering instructions provided in an
engineering change request document
2003-02-02
Standby gas treatment system inoperable due to wrong damper position
in operating procedure
2003-02-01
Inadequate corrective action for the failure of standby gas treatment
system fan EF-1-8
2002-08-03
Failure to maintain surveillance test procedure 607.4.004, Containment
Spray/Emergency Service Water System Pump Operability
2002-08-06
Inadequate Procedural Guidance & Personnel Performance Resulting in a
Plant Event
2003-03-03
Failure to implement procedure for relocation of primary whole-body
dosimetry
2002-08-07
Ineffective Resolution of Identified Problems with Personnel Response to
Alarming SRDs
2001-13-01
Band Clamps of Different Design Found on Several HCUs Without
Engineering Evaluation
Audits and Self-Assessments:
Emergency Preparedness, 50.54T, Meteorology, June 2003
FASA, Human Performance Baseline Assessment, February 2002
FASA, Clearance and Tagging, March 2003
FASA, Fire Protection, December 2001
FASA, Plant Operations Review Committee Program, September 2003
FASA, Problem Identification and Resolution, April 2004
Regulatory Assurance Corrective Action Monthly Report, January 2004
Regulatory Assurance Corrective Action Monthly Report, February 2004
Regulatory Assurance Corrective Action Monthly Report, December 2003
NOSA OC-03-03, April 7-11, 2003.
Effectiveness review - CAP2003-0716
Effectiveness review - CAP2003-0638
Effectiveness review - CAP2001-0251
FASA, Corrective action effectiveness - Engineering
NOSA-OC-03-05, Design Engineering, 8/10/03
Focused Area Self-Assessment Report, AR158759158759 Document Management and Administrative
Control of Drawings, 5/14/03
Regulatory Assurance Corrective Action Process Monthly Report, December 2003
Regulatory Assurance Corrective Action Process Monthly Report, February 2004
Regulatory Assurance Corrective Action Process Monthly Report, January 2004
FASA, System Engineering Performance Monitoring, July 2003
FASA, Human Error Prevention, March 2003
NOSA-OC-03-06, Health Physics & Radiation Protection Audit
NOSA-OYS-04-01, Maintenance Functional Area, March 2004
2002-1424, Effectiveness Review - Electronic Dosimetry Usage
2003-0204, Effectiveness Review - High Radiation Area Boundary Control
Condition Action Program Reports (* denotes a CAP generated as a result of this
inspection):
DR 94-6-24
O2000-1212
O2001-1379
O2001-1843
O2002-0099
O2002-0108
O2002-0255
O2002-0447
O2002-0449
O2002-0459
O2002-0545
O2002-0565
O2002-0955
O2002-1089
O2002-1184
O2002-1186
O2002-1551
O2002-1589
O2003-1000
O2003-1814
O2003-0889
O2003-1962
O2003-2257
O2004-0311
O2004-0555
O2003-0011
O2003-0334
O2004-0040
O2002-0157
O2004-0618
O2003-0716
O2002-1656
O2004-0412
O2002-1616
O2002-1710
O2002-1891
O2002-1945
O2003-0017
O2003-0030
O2003-0055
O2003-0058
O2003-0059
O2003-0063
O2003-0175
O2003-0254
O2003-0337
O2003-0399
O2003-0435
O2003-0441
O2003-0473
O2003-0546
O2003-0652
O2003-0654
O2003-0670
O2003-0700
O2003-0773
O2003-1097
O2003-1157
O2003-1158
O2003-1166
O2003-1167
O2003-1168
O2003-1169
O2003-1170
O2003-1171
O2003-1173
O2003-1174
O2003-1176
O2003-1179
O2003-1203
O2003-1204
O2003-1323
O2003-1362
O2003-1381
O2003-1465
O2003-1579
O2003-1622
O2003-1674
O2003-1765
O2003-2015
O1998-0823
O2000-0261
O2000-0617
O2000-0691
O2001-0087
O2001-0084
O2001-1741
O2001-1908
O2002-0006
O2002-0057
O2002-0059
O2002-0452
O2002-0542
O2002-0640
O2002-0711
O2002-0797
O2002-0801
O2002-0850
O2002-0980
O2002-1059
O2002-1343
O2002-1538
O2002-1579
O2002-1643
O2002-1709
O2002-1951
O2002-1977
O2002-1992
O2002-2013
O2003-0184
O2003-0273
O2003-0413
O2003-0469
O2003-0616
O2003-0664
O2003-0686
O2003-0868
O2003-0776
O2003-0881
O2003-1056
O2003-1087
O2003-1131
O2003-1150
O2003-1189
O2003-1260
O2003-1270
O2003-1282
O2003-1592
O2003-1593
O2003-1595
O2003-1603
O2003-1604
O2003-1606
O2003-1607
O2003-2050
O2003-2094
O2003-2101
O2003-2104
O2003-2212
O2003-2230
O2003-2370
O2003-2596
O2004-0122
O2004-0123
O2004-0162
O2004-0165
O2004-0477
O2004-0492
O2004-0616
O2004-0620
O2004-1023*
O2004-1231*
O2004-1240*
O1998-1541
O2000-1839
O2001-1773
O2001-0025
O2001-1320
O2004-0356
O2004-0620
O2004-0220
O2003-1573
O2003-1286
O2004-0023
O2003-2663
O2003-2496
O2003-2279
O2003-1843
O2003-2172
O2002-1689
O2003-0290
O2002-1557
O2003-2064
O2003-1868
O2003-1746
O2003-1131
O2003-0919
O2003-0809
O2003-0540
O2003-0541
O2003-0393
O2003-1401
O2003-0202
O2003-0073
O2002-1678
O2002-1775
O2003-0518
O2003-0796
O2003-0796
O2002-0542
O2003-1209
O2003-0753
O2003-1308
O2003-1331
O2002-1424
O2003-0204
2002-0711
O2003-1581
O2003-1365
O2002-1059
O2003-1621
O2004-1253*
O2003-1616
O2003-1622
O2004-1111*
O2002-1808
O2002-0201
O2002-1919
O2002-1363
O2004-1247*
O2003-0884
O2003-1161
O2003-2534
O2002-1505
O2002-1365
O2001-1155
O2002-1626
O2004-0858
O2003-0970
O2003-1052
O2003-1172
O2002-1662
O2003-2577
O2004-0508
O2002-1367
O2003-0310
O2002-1951
O2002-1992
O2002-0249
O2002-1015
O2003-2557
O2003-1637
O2003-1923
O2003-1591
O2003-1414
O2003-2237
O2003-1715
O2003-0266
O2001-1447
O2002-1788
O2004-0296
O2003-0627
O2001-0584
O2001-0524
O2002-0249
O2001-0556
O2003-2577
O2003-0912
O2002-1578
O2002-1451
O2003-1715
O2003-0566
O2002-1798
O2003-0566
O2003-0483
O2002-1798
O2003-1024
O2003-1021
O2002-1676
O2002-1680
O2004-1242*
O2004-1203*
O2004-1219*
O2004-1230*
O2004-1136*
O2004-1036*
O2004-1037*
O2004-1124*
O2004-1205*
O2003-0030
O2003-1552
Action Requests:
A2073528
A2049122
A2031368
A2046137
A0706057
A2058772
A0156470
A2047330
A0775402
A0031808
A2014242
A2068349
A0703272
A2042874
A0700109
A2008192
A2013213
A2012559
A2021781
A2022260
A2023446
A2030944
A2046024
A2071201
A2063490
A2064754
A2069334
A2069867
A2073413
A2076849
A2077924
A2078783
A2079850
A2084493
A2011234
A2084792
A2085472
A2086655
A2086657
A2045009
A2029380
A2031091
A2040692
A2049913
A2034412
A2042215
A2045531
A2057287
A2048084
A2034841
A2068100
A2045756
A2047330
A2051820
A2057830
A2059656
A2060261
A2062379
A2064992
A2044853
A2079519
A2038132
A2061761
A2072346
A2070845
A2025957
A2031857
A2036807
A2036850
A2045092
A2050630
A2051865
A2055989
A2057186
A2061416
A2065866
A2077583
A2080377
A2086314
A2086927
A2078161
A2059150
A2011234
A2034841
A2038100
A2044413
A2035693
A2029911
A2046276
A2062455
A2045487
A2046161
A2034624
A2030146
A2046161
A2077258
A2036165
A2084438
A2069321
A2079459
A0707408
A2061943
A2057596
A2057874-E27
A2036044-E23
A2061522-E01
A2032586
A2021319
A2021452
A2030066
OPERABILITY EVALUATIONS (for CAPs):
O2002-0233, Aging Agastat Time Delay Relays for the Isolation Condenser
O2002-0716, EDG-2 Fuel Oil Return Sight Glass Half-Full
O2002-1808, Emergency Service Water Flow Outside IST Action Limits
O2003-0317, Emergency Service Water Operability During Low Intake Levels
O2003-1865/2225, Water and Sediment in Main Fuel Oil Tank
O2003-2225, EDG Fuel Oil Tank
O2003-2399, Aggregate Review of Emergency Service Water CAPs
O2004-0110, Operability Determination of Emergency Service Water Pump 52B
O2002-0157, 480V undervoltage trip devices
O2003-2017, Battery charger C1
O2002-1059, Service Water Piping between RBCCW Hx and Seal Well
O2002-1808, Emergency Service Water Pump Performance
O2002-1551,Void between Reactor and Turbine Building walls below 23 floor affects Appendix
R separation
DRAWINGS:
GU3C-733-11-010, 120V AC vital power system, Revision 5
913E911, RPS MG set control, Revision 0
NQZ-0001, Vacuum Breaker position indicating system, Revision 3
GE148F740, Containment Spray System, Revision 43
BR-2005, Emergency Service Water System (page 4 of 6), Revision 73
GU-3E-243-21-1000, Drywell and Torus vacuum relief system flow diagram, Revision 28
3E-862-21-1000, Emergency Diesel Generator Diesel Fuel Oil Storage and Transfer System
Flow Diagram, Revision 21
3E-861-21-1001,Emergency Diesel Generator Water Cooling System Flow Diagram, Rev. 10
3E-861-21-1002, Emergency Diesel Generator Lube Oil System Flow Diagram, Revision 11
3E-861-21-1000, Emergency Diesel Generator Air Cooling System Flow Diagram, Revision 11
BR2013, Sheet 3, Instrument (Control) Air System Flow Diagram, Revision 61
BR2013, Sheet 4, Instrument (Control) Air System Flow Diagram, Revision 55
BR2013, Sheet 5, Instrument (Control) Air System Flow Diagram, Revision 59
BR2013, Sheet 6, Instrument (Control) Air System Flow Diagram, Revision 69
BR2013, Sheet 7, Instrument (Control) Air System Flow Diagram, Revision 59
BR2013, Sheet 8, Instrument (Control) Air System Flow Diagram, Revision 59
BR2013, Sheet 9, Instrument (Control) Air System Flow Diagram (Fluid Details New Radwaste),
Revision 57
BR2013, Sheet 10, Instrument (Control) Air System Flow Diagram (Fluid Details New
Radwaste), Revision 52
BR3000, Electrical Power System Key One Line Diagram, Revision 6
BR3001, Sheet 1, Revision 6, Plant Electrical Generation Main One Line Diagram Auxiliary
Startup & Main Xfmrs, SBO Xfmr and Main Generator
BR3001, Sheet 2, Revision 4, Plant Electrical Generation Main One Line Diagram Auxiliary
Startup & Main Xfmrs, SBO Xfmr and Main Generator
BR3001A, Revision 3, 4160V System One Line Diagram, 4160 Swgr. Bus 1A
BR3001B, Revision 8, 4160V System One Line Diagram, 4160 Swgr. Bus 1B & Dilution Plant
BR3001C, Revision 0, 4160V System One Line Diagram, 4160 Emergency Swgr. Bus 1C & 1D
BR3002, Sheet 1 of 4, Revision 8, 480V System One Line Diagram, 460V Unit Substation 1A1
& 1B1
BR3002, Sheet 2 of 4, Revision 4, 480V System One Line Diagram, 480V Unit Substation 1A2
& 1B2
BR3002, Sheet 3 of 4, Revision 4, 480V System One Line Diagram, 460V Unit Substation 1A3
& 1B3
BR3002, Sheet 4 of 4, Revision 4, 480V (JCP&L) Non-Vital Power One Line Diagram, 480V
Unit Substation 1C1
4053, Sheet 13, Reactor Building First Floor At El. 236" Plan And Details
4093-6, Sheet 2; Turbine Building Floor Plan @ El. 270" & 36-0", Beam & Slab Schedules
2003-025
2003-020
2004-026
2002-041
2003-001
2002-012
2003-018
2003-019
2003-020
2004-021
2004-025
WORK ORDERS:
M2059656 02
M2059656 03
M2059656 04
M2059656 06
C2003513
C2004238
C2002648
M2059656 07
C0519083
C0000403
C2006219
A2082612
C2006233
C2006368
C2006429
A2012559
C2006134
C2006567
DOSIMETRY INVESTIGATION REPORTS:
DIR 03-017
DIR 03-027
DIR 03-049
DIR 03-098
DIR 03-033
DIR 03-052
DIR 03-100
MISCELLANEOUS DOCUMENTS:
ECR No.OC-01-01193-001, Install New APRM Cards for Increased Core Flow
Applicability to Oyster Creek of Generic Letter 97-17, Cracking of Vertical Welds in the Core
Shroud and Degraded Repair
Oyster Creek Nuclear Generating Station Core Shroud Repair - Design Report, Volume 1 of 2,
Revision 1
Fire Hazards Analysis Report, Revision 12
Health Report for System 811, Fire Protection Water System
Health Report for System 532, Emergency Service Water System
Health Report for System 241, Containment Spray System
Health Report for System 741, Emergency Diesel Generators (Electrical)
Training Lesson Plan for Containment Spray/Emergency Service Water Systems
Training Lesson Plan for Fire Protection System
Applicability to Oyster Creek of Generic Letter 91-06, Resolution of Generic issue A-30,
Adequacy of safety related DC power supplies, pursuant to 10 CFR 50.54(f)
Fire Hazards Analysis Report, Revision 12
NEDC-33058P, DRF 0000-0000-7015, Class III, July 2002; Increased Core Flow Analysis For
Oyster Creek Generating Station
Calculation C-1302-223-E170-043, Revision 0, 3/1/01
TDR-914, Revision 1, 10/13/89; Evaluation of Instrument Air Loss To System Air Operated
Valves and Dampers At Oyster Creek
Calculation C-1302-826-5360-007, Revision 0, 12/5/89; OC Control Room/Cable Room
Temperature
2.4.002, completed 8/22/03
2.4.002, completed 8/23/03
2.4.002, completed 8/22/03
603.3.004, completed 10/25/1998
603.3.004, completed 11/04/1998
603.3.004, completed 11/09/2000
603.3.004, completed 10/23/2002
607.4.004, completed 11/13/2002
Engineering Evaluation No.125, Install Encapsulation Device On Piping Upstream Of V-6-0024
Plant Health Committee System Presentation P851/852 Service & Instrument Air, March 04
Quarterly Ship System Report, P851/852 Service & Instrument Air; March 1, 04
Oyster Creek Nuclear Generating Station, Fire Hazards Analysis Report, Document 990-1746,
Revision 12
System/Component Walkdown Checklist - Service & Instrument Air System; 5/7/03, 11:00 AM
50.59 Review for ECR 03-00028 and ECR 03-00155; Anti-Siphon Holes Added To The SFP
Cooling System Return Lines, 2/25/03
ECR OC 03-00851
NRC Safety Evaluation Dated January 25, 1990; Exemption From Certain Technical
Requirements Contained In Section III. G Of Appendix R to 10 CFR 50
LER 02-003; Insufficient Appendix R Separation Criteria Due to Sand Erosion; dated 12/6/02
LER 02-003-01; Insufficient Appendix R Separation Criteria Due to Sand Void; dated 9/19/03
LER 02-003-02; Insufficient Appendix R Separation Criteria Due to Sand Void; dated 4/2/04
50.59 Review for Modification ECR 03-00851
SEN 242
Licensee Event Report 50-0219/2002-01
Licensee Event Report 50-0219/2002-02
Nuclear General Employee Training, Revision 28
OC Topical Report 140, "Emergency Service Water & Service Water Systems Piping Plan
Drawing 2005, sheet 2, "Service Water System - Reactor & Turbine Bldg"
Drawing 2006, sheets 1-3, "Reactor Bldg Closed Cooling Water System (RBCCW)"
Drawing GE 729E183, sheets 1-5, "Auto Depressurization System"
RBCCW System Walkdown Checklists, dated 3/22/02, 6/19/02, 10/7/02, 3/28/03, 6/20/03,
9/29/03, 10/22/03, 12/31/03, and 3/15/04
System Health Reports for Service Water System and RBCCW
Training Lesson Plans for Service Water System and RBCCW
Maintenance Rule Performance Documents for Reactor Protection, Main Steam, Feedwater,
Condensate-2, and Isolation Condenser, and Fire Protection Water Systems
LIST OF ACRONYMS USED
AmerGen
AmerGen Energy Company, LLC
Action Request
Corrective Action Program Report
CFR
Code of Federal Regulations
Emergency Service Water
Focused Area Self Assessment
Instrument Air
IMC
Inspection Manual Chapter
Individual Plant Examination
Non-Cited Violation
NOS
Nuclear Oversight
NRC
Nuclear Regulatory Commission
Operating Experience
Problem Identification and Resolution
Reactor Building Component Cooling Water
Revised Oversight Program
Significant Determination Process
Vac
Volts Alternating Current
Vdc
Volts Direct Current