ML050540469

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Dynamic Web Page Printout Dated 02/10/2004, ROP PIM Report - All Sites - Event Dates: 01/01/2003 - 02/10/2004
ML050540469
Person / Time
Site: Dresden, Oconee, Perry, Wolf Creek, Point Beach, Hope Creek, Arkansas Nuclear, Columbia, Surry, North Anna, Diablo Canyon, Callaway, Waterford, Cook  Constellation icon.png
Issue date: 02/10/2004
From:
Office of Nuclear Reactor Regulation
To:
References
FOIA/PA-2004-0277
Download: ML050540469 (5)
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Text

Dynamic Web Site Page 1 of 5 Dynamic Web Page ROP PIMI Report - All Sites - Event Dates: 01/01/2003 - 02/10/2004 By Types, Cornerstones, Event Dates, Sites Significance: "TBD" There are 18 Qpen/Cosed PlMs selected en thbis web page generated on 02loo/2004 for all regions.

Appamnt Violation - TlBD 6

NonCited Vioation -TBD I

Unresolved tem -TBD 11 Cross Cutting Areas:

  • SCWE - Safety Conscious }fork Environment
  • HP - Human Performance
  • SCWE: N
  • HP: N
  • PIR: N Docket/Status: 05000266 (0), 05000301 (0)

Open: 2003007 ID CFR 50.54, 10 CFR 50.47 apparent violation for failure to maintain a standard scheme of emergency action levels The inspectors identified an apparent violation of 10 CFR 50.54(q), associated with emergency planning standard 10 CFR 50.47(b)(4), which will be subject to the NRC traditional enforcement process not the revised Reactor Oversight Process. Specifically, the licensee failed to maintain a standard scheme of emergency action levels (EALs). Eight EALs were changed in 1998 and 1999. The changes decreased the effectiveness of the Emergency Plan in that emergency conditions that would have resulted in classifications at the General Emergency (GE), Alert, and Notification of Unusual Event (NOUE) levels would result in a lesser classification under the current EAL scheme. Approval of the NRC was not obtained prior to the changes being made. Since the identification of the issue by the inspectors, the licensee has revised the eight EALs to be equivalent with those approved by the NRC in 1984.

Emergency Preparedness 08/22/2003 ANO TBD

  • SCWE: N
  • HP: N
  • PIR: N Docket/Status: 05000313 (0), 05000368 (0)

Open: 2003011 A

0311 -01 FAILURE TO MEET THE ALERT NOTIFICATION SYSTEM DESIGN CRITERIA TBD. The inspector identified a violation of 10 CFR 50.54(q) having a potential safety significance greater than very low significance because the licensee failed to follow the emergency plan requirement to establish a means to notify members of the public in the emergency planning zone. Between September 1999 and April 2003, a small percentage of residences in the licensee's plume exposure emergency planning zone would not have received an emergency alerting signal in the event of an emergency at the Arkansas Nuclear One facility. The finding had greater than minor significance because the condition resulted in a loss of alert notification capability to a small percentage of the emergency planning zone population, and if left uncorrected the condition would have continued to degrade. Using the Emergency Preparedness Significance Determination Process the finding was preliminarily determined to have low to moderate safety significance (White) because it was a violation of 10 CFR 50.54(q) and represented a degradation of the risk-significant planning standard 10 CFR 50.47(b)(5) function.

Mitigating Systems 01/23/2004 OCO TBD

  • SCWE: N
  • HP: N
  • PIR: N Docket/Status: 05000269 (0), 05000270 (0), 05000287 (0)

Open: 2004007 Failure to Obtain Prior NRC Approval to a Change to the Facility Involving Unreviewed Safety Questions on High Energy Line Break Analysis The inspectors identified an apparent violation of 10 CFR 50.59 (a)(l) (1999 version of 10 CFR) which states, in part, that the licensee may make changes in the facility as described in the safety analysis report without prior Commission approval, provided the proposed change does not involve an unreviewed safety question (USQ). 10 CFR 50.59 (a)(2) states, in part, that a proposed change involves an USQ if the probability of occurrence or malfunction of equipment important to safety previously evaluated in the safety analysis report may be increased, or if it may create an accident different from any previously evaluated. On May 17, 2001, the licensee made a change to the facility, as described in the Updated Final Safety Analysis Report, Section 3.6.1.3, associated with the High Energy Line Break (HELB) analysis, which involved unreviewed safety questions, and failed to obtain prior NRC

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Dynamic Web Site Page 2 of f approval. The UFSAR Section was changed to increase the maximum initiation time following HELB of Emergency Feedwater from 15 to 30 minutes and of High Pressure Injection from I hour to 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> (based on referenced reports and analysis). The analysis discussed an increased cycling of pressurizer Safety Relief Valves on steam and water, boiler condenser mode of decay heat removal, and an unapproved computer code for application to HELB, but failed to recognize that such changes may increase the probability of occurrence or the consequences of a malfunction of equipment important to safety or may create an accident different from any previously evaluated. In addition, the change resulted in more than a minimal increase in risk. (Section 40A5)

Mitigating Systems 12/31/2003 PERR TBD

  • SCWE: N
  • HP: N
  • PIR: Y Docket/Status: 05000440 (0)

Open: 2003010 INADEQUATE LPCS/RHR 'A' FILL AND VENT PROCEDURES RESULTS IN SYSTEM INOPERABILITY AFTER LOSS OF OFFSITE POWER An apparent self-revealed violation of Technical Specification 5.4 occurred when the waterleg pump for low pressure core spray (LPCS) and residual heat removal (RHR) 'A' became air bound following a loss of offsite power. Subsequent investigation revealed that the procedures for venting these systems did not include the high point vent valve on the discharge of the pump, thus allowing gas to accumulate in a vertical section of system piping. When the waterleg pump lost power on August 14, 2003, the accumulated gas expanded and caused voiding of the pump. As a result, both LPCS and RHR 'A' were rendered inoperable. The NRC assessed this finding through Phase 3 of the Significance Determination Process and made a preliminary determination that it is an issue with low to moderate safety significance.

Mitigating Systems 12/31/2003 WAT TBD

  • SCWE: N
  • HP: Y l *PIR: N Docket/Status: 05000382 (0)

Open: 2003007 Failure to establish appropriate instructions and implement those instructions A self-revealing apparent violation of 10 CFR Part 50, Appendix B, Criterion V, "Instructions, Procedures, and Drawings," was identified for the failure to establish appropriate instructions and accomplish those instructions for installation of a fuel line for Train A emergency diesel generator in May 2003. This failure resulted in uneven and excessive scoring of the tubing that ultimately led to a complete 360 degree failure of the fuel supply line on September 29, 2003, during a monthly surveillance test. This finding is unresolved pending completion of a significance determination. The finding was greater than minor because it directly impacted the availability and reliability of an emergency diesel generator which is used to mitigate the loss of AC power to the respective safety related bus. The finding was determined to have a potential safety significance greater than very low significance because the failure resulted in an actual loss of the safety function of the Train A emergency diesel generator for an extended period of time.

Public Radiation Safety l 10/08/2003 COOK TBD

  • SCWE: N
  • HP: N
  • PIR: N Docket/Status: 05000315 (0), 05000316 (0)

Open: 2003016 Failure to Prepare a Shipment of Radioactive Waste to Satisfy Department of Transportation External Package Radiation Level Limits A self-revealed finding preliminarily assessed to be greater than Green and an associated apparent violation (AV) were identified for the failure to prepare a package of radioactive material for shipment, so that under conditions normally incident to transportation, the radiation level does not exceed 200 millirem/hour at any point on the external surface of the package. Package surface radiation levels in excess of 200 millirem/hour were identified by a waste processing contractor upon receipt of the shipment from the licensee. The finding was more than minor because it was associated with the "Program and Process" attribute of the Public Radiation Safety Cornerstone, and affected the cornerstone objective of ensuring adequate protection of public health and safety from exposure to radioactive materials released into the public domain. Also, the issue involved an occurrence in the licensee's radioactive material transportation program that was contrary to NRC and Department of Transportation (DOT) regulations. The finding was determined preliminarily to be of low to moderate safety significance because the transportation problem involved an external package radiation level that exceeded limits by 25 percent and because the area of elevated radiation on the package was determined to be accessible to a member of the public during conditions normally incident to transportation. To address this issue, the licensee planned to revise procedures to require load plans and to specify which survey instrumentation is to be used for package surveys, and to provide training to its staff involved in radioactive material shipments.

NonCited Violation Mitigating Systems l1231/2003 DIAB TBD

  • SCWE: N
  • HP: N
  • PIR: Y Docket/Status: 05000275 (C), 05000323 (C)

Open: 2003008 Failure to Adequately Train Operations Responders in Support of the Fire Brigade The inspectors identified a violation of Technical Specification 5.4.1.d which requires written procedures be established, implemented and maintained http://nrrIO.nrc.gov/rps/dyn/rop-irl.cfm 02/1012004

IDynamic Web Site Page 3 of 5 covering the Fire Protection Program implementation. Specifically, PG&E failed to adequately establish and implement procedural changes that provided for senior control operators, licensed control operators and non-licensed, level 8 nuclear operators to serve in the operator responder position. The inspectors noted that the applicable attachment to the procedure for conduct of the operations response position was not established until after training had been provided on implementing the procedure. Operations responders supporting the fire brigades exhibited a knowledge weakness in activities such as communications with the control room, manual actuation of fire suppression equipment, and providing iniformation to the fire brigade regarding safe shutdown equipment. The finding is unresolved pending completion of a significance determination. The finding is greater than minor because it affects the mitigating system cornerstone objective by degrading fire brigade effectiveness, which is a fire protection defense-in-depth element.

Unresolved item Barrier Integrity 10/2112003 CALL TBD

  • SCWE: N
  • HP: N
  • PIR: N Docket/Status: 05000483 (0)

Open: 2003007 Inadequate Alarm Response Procedure for Smoke in the Control Room Ventilation Supply Duct The alarm response procedure for responding to smoke in the control room outside supply duct was inadequate because it did not direct operators to isolate outside air makeup upon receipt of the alarm. This alarm does not cause an automatic isolation of the control room, so operators must recognize the condition and take manual action to prevent losing control room habitability. Failure to have a procedure, required by Technical Specification 5.4.1.a and Regulatory Guide 1.33, that provided appropriate response actions for abnormal or alarm conditions was a violation. This issue was entered into the licensee's corrective action program under Callaway Action Request 200306977. This finding is unresolved pending completion of a significance determination. This issue was more than minor because failure to isolate the control room ventilation could lead to unnecessary evacuation, which would result in a plant transient and disabling much of the mitigation equipment that would otherwise be available. This issue is being treated as an unresolved, item pending completion of a significance determination.,

Initiating Events 09/27/2003 HOPE TBD

  • SCWE: N
  • HP: N
  • PIR: N Docket/Status: 05000354 (0)

Open: 2003005 Discussed: 2003006 MAINTENANCE EFFECTIVENESS ON 'A" STATION SERVICE WATER SYSTEM TRAVELING SCREEN A self-revealing finding occurred when the A SSWS traveling screen failed and PSEG determined a contributing cause was improper cutting of a key without procedure guidance. The inspectors identified an additional problem that contributed to the failure when applicable maintenance procedures were not used to set traveling chain tension and screen level. This performance issue was determined to have potential safety significance greater than very low safety significance based on preliminary risk assessments that considered the associated pump unavailable while the traveling screen was inoperable.

Initiating Events 05/23/2003 NA TBD

  • SCWE: N
  • HP: N
  • PIR: N Docket/Status: 05000338 (0), 05000339 (0)

Open: 2003006 Alternate Shutdown Panel Ventilation System Not Independent from Impacts of a Main Control Room Fire The shared ventilation system between the Main Control Room (MCR) and the Unit I and Unit 2 Emergency Switchgear and Relay Rooms (ESGRs) do not have adequate separation, isolation, or barriers to prevent smoke and toxic gases from being transported to the ESGRs during a fire in the MCR. The alternative shutdown capability for an MCR fire is located at the auxiliary shutdown panels in each unit's ESGR, respectively. This finding is unresolved pending completion of a significance determination. The finding is greater than minor because it affects the mitigating systems cornerstone objectives. The finding has potential safety significance greater than very low safety significance because operator inability to safely man the auxiliary shutdown panels could result in failure of the specified alternative shutdown strategy.

Miscellaneous 1 03/31/2003 DRES TBD

  • SCWE: N
  • HP: N
  • PIR: N Docket/Status:, 05000237 (0), 05000249 (0)

Open: 2003002 Adequacy of Site Welding Program Mitigating Systems 12/31/2003 CALL TBD

  • SCWE: N
  • HP: N
  • P]R: N Docket/Status: 05000483 (0)

Open: 200300 http://nrrlO.nrc.gov/rps/dyn/rop-irl.cfm 02/10/2004

I Dynamic Web Site Page 4 of 5 Failure to maintain an Emergency Operating Procedure consistent with the accident analysis.

A finding was identified concerning a sequencing error in Emergency Operating Procedure E-3, 'Steam Generator Tube Rupture' that could result in increased post-accident public radiation dose. The sequence error delayed termination of safety injection during simulator exercises. The prolonged accident recovery time increased the postulated radiological source term released during the accident. The failure to maintain the EOP consistent with the accident analysis was an apparent violation of Technical Specification 5.4, 'Procedures." This issue was entered into the licensee's corrective action program as CAR 200304922. This finding is unresolved pending completion of a significance determination. This issue was more than minor because the EOP quality attribute of the barrier integrity cornerstone is affected by the procedural error.

Mitigating Systems 12/31/2003 WC TBD

  • SCWE: N
  • HP: N
  • PIR: N Docket/Status: 05000482 (0)

Open: 200300 Fire Barrier in the Main Steam Enclosure Missing The inspectors identified a violation of License Condition 2.C(5)(a) of the Wolf Creek Generating Station Facility Operating License having potential safety significance greater than very low significance because approximately 20 inches of fire barrier between the main steam enclosure and auxiliary feedwater system flow control valve rooms was missing. The finding is unresolved pending completion of a significance determination. The finding is greater than minor because it is associated with a degraded Fire Protection fire barrier and affected the Reactor Safety Mitigating System cornerstone. The finding was determined to have potential safety significance greater than very low significance because all the main steam atmospheric relief and auxiliary feedwater system flow control valves could be affected by a fire in either area.

Mitigating Systems 05/23/2003 NA TBD

  • SCWE: N
  • HP: N
  • PIR: N Docket/Status: 05000339 (0)

Open: 2003006 Fire Response Procedure 2-FCA-2 Not Adequate To Assure Safe Shutdown Of Unit 2 The safe shutdown strategy and related fire response procedures may be inadequate to assure a safe shutdown of the Unit 2 reactor for a fire in Emergency Switchgear and Relay Room (ESGR) No. 2. The licensee's fire response procedures may not preclude plant damage and may prescribe operator actions in the Cable Vault and Tunnel that are not independent from the effects of an ESGR No. 2 fire. This finding is unresolved pending completion of a significance determination. The finding is greater than minor because it affects the initiating event and mitigating systems cornerstone objectives. Also, the finding has potential safety significance greater than very low safety significance because in some scenarios, these deficiencies could lead to reactor coolant pump seal package leakage and failure of the specified alternative shutdown strategy.

Mitigating Systems 04/22/2003 WNP TBD

  • SCWE: N
  • HP: N
  • PIR: N Docket/Status:.05000397 (0)

Open: 2003002 Failure to have adequate alternative shutdown procedures The inspectors identified a violation of Technical Specification 5.4.1.d (inadequate procedure) because Procedure ABN-CR-EVAC, "Control Room Evacuation and Remote Cooldown," failed to provide adequate post-fire direction to: (I) assure suppression pool temperatures did not increase above residual heat removal pump temperature limits following depressurization; and (2) assure adequate core cooling with one safety relief valve stuck open. This finding is unresolved pending completion of a significance determination. This finding is greater than minor because it impacts the mitigating systems cornerstone and affects the ability of the low pressure coolant injection system to provide adequate core cooling to prevent core damage. This finding was determined to have potential safety significance greater than very low significance because of the lack of credited systems to mitigate the effects of a control room fire.

Mitigating Systems 04/22/2003 WNP TBD

  • SCWE: N
  • HP: N
  • PIR: Y Docket/Status:, 05000397 (0)

Open: 2003002 Inadequate corrective actions to address water hammer concern The inspectors identified a violation of License Condition 2.C(14) for the failure to take appropriate corrective measures to address a condition adverse to quality affecting the low pressure coolant injection system. During a control room fire, the system has been vulnerable to a water hammer since at least 1997 due to a leaking check valve in Train B of the residual heat removal system. The licensee took over five years to identify the condition and failed to specify appropriate corrective measures to promptly fix the condition. This finding is unresolved pending completion of a significance determination. This finding is greater than minor because it impacts the mitigating systems cornerstone and affects the ability of the low pressure coolant injection system to provide http://nrrlO.nrc.gov/rps/dyn/ropirl.cfm 02/10/2004

Dynamic W& Site Page 5 of 5 adequate core cooling to prevent core damage. This finding was determined to have potential safety significance greater than very low significance because of the lack of credited systems to mitigate the effects of a control room fire Mitigating Systems 02/1412003 SUR TBD lSCNVE: N

  • HP: N
  • PIR: N Docket/Status: 05000280 (0)

Open: 2003007 Fire Response Procedures l-FCA-4.00 And 0-FCA-14.00 Not Adequate To Assure Safe Shutdown Of Unit I The safe shutdown strategy and related fire response procedures may be inadequate to assure a safe shutdown of the Unit I reactor for a fire in Emergency Switchgear and Relay Room (ESGR) Number 1. The licensee's fire response procedures may not preclude plant damage, may fail to prevent potential spurious operations and may require the operator to enter the affected fire area to perform directed actions. This finding is unresolved pending completion of a significance determination. The finding is greater than minor because it was associated with the ability to achieve a safe shutdown of the Unit I reactor following a fire in ESGR No. I and affects the initiating event and mitigating systems cornerstone objectives. Also, the finding has potential safety significance greater than very low, safety significance because RCP seal package failure could cause a seal loss-of-coolant accident and failure of the specified alternative shutdown strategy.

Mitigating Systems 02/14/2003 SUR TBD

  • SCWE: N
  • HP: N
  • PIR: N Docket/Status: 05000280 (0)

Open: 2003007 Fire Response Procedures l-FCA-3.00 And 0-FCA-14.00 Not Adequate To Assure Safe Shutdown Of Unit I The safe shutdown strategy and related fire response procedures may be inadequate to assure a safe shutdown of the Unit 1 reactor for a fire in the Unit I cable vault and cable tunnel. The licensee's fire response procedures may not preclude plant damage, may fail to prevent potential spurious operations and may require the operator to enter the affected fire area to perform directed actions. This finding is unresolved pending completion of a significance determination. The finding is greater than minor because it was associated with the ability to achieve a safe shutdown of the Unit I reactor following a fire in the Unit I cable vault and cable tunnel and affects the initiating event and mitigating systems cornerstone objectives. Also, the finding has potential safety significance greater than very low, safety significance because RCP seal package failure could cause a seal loss-of-coolant accident and failure of the specified alternative shutdown strategy.

http://nrrlO.nrc.gov/rps/dyn/ropjrl.cfm 02/1012004

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