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Category:Licensee Event Report (LER)
MONTHYEAR05000285/LER-1917-001, Regarding Unprotected Vital Area Barrier Due to Maintenance2017-05-11011 May 2017 Regarding Unprotected Vital Area Barrier Due to Maintenance 05000285/LER-2016-003, Regarding Unplanned Turbine Trip During DCS Modification Due to Failure to Identify and Disable the Transmitter Deviation Based Trip2016-08-22022 August 2016 Regarding Unplanned Turbine Trip During DCS Modification Due to Failure to Identify and Disable the Transmitter Deviation Based Trip 05000285/LER-2016-002, Regarding Unanalyzed Condition Shutdown Heat Exchanger Isolations2016-07-0707 July 2016 Regarding Unanalyzed Condition Shutdown Heat Exchanger Isolations 05000285/LER-2016-001, Regarding Technical Specification Violation Due to Installation of an Unqualified Part in a Radiation Monitor2016-04-0808 April 2016 Regarding Technical Specification Violation Due to Installation of an Unqualified Part in a Radiation Monitor 05000285/LER-2015-006, Regarding Unanalyzed Fire Vulnerability Due to Inadequate Design2015-12-11011 December 2015 Regarding Unanalyzed Fire Vulnerability Due to Inadequate Design 05000285/LER-2015-005, Regarding Reactor Coolant Leak at Reactor Coolant Pump Seal Due to Cyclic Fatigue2015-09-17017 September 2015 Regarding Reactor Coolant Leak at Reactor Coolant Pump Seal Due to Cyclic Fatigue 05000285/LER-2015-004, Regarding Inoperability of Auxiliary Feedwater Trains Due to Failure of Steam Generator Isolation Valve2015-08-0303 August 2015 Regarding Inoperability of Auxiliary Feedwater Trains Due to Failure of Steam Generator Isolation Valve 05000285/LER-2015-003, Regarding Containment Spray Inoperable Due to Original Design Error2015-06-15015 June 2015 Regarding Containment Spray Inoperable Due to Original Design Error 05000285/LER-2015-002, Regarding Inoperable Auxiliary Feedwater System Due to Inadequate Procedure Change2015-05-28028 May 2015 Regarding Inoperable Auxiliary Feedwater System Due to Inadequate Procedure Change 05000285/LER-2015-001, Regarding Inadequate Design of High Energy Line Break Barriers2015-03-27027 March 2015 Regarding Inadequate Design of High Energy Line Break Barriers 05000285/LER-2014-007, Regarding Plant Trip Due to Moisture Intrusion Into a Transformer Control Cabinet2015-02-13013 February 2015 Regarding Plant Trip Due to Moisture Intrusion Into a Transformer Control Cabinet 05000285/LER-2013-014-02, Regarding Unqualified Components Used in Safety System Control Circuit2015-01-29029 January 2015 Regarding Unqualified Components Used in Safety System Control Circuit 05000285/LER-2014-006, Regarding Inoperability of Radiation Monitors Due to an Error in Technical Specifications2014-11-12012 November 2014 Regarding Inoperability of Radiation Monitors Due to an Error in Technical Specifications 05000285/LER-2014-005, Regarding Technical Specification Violation of Containment Integrity2014-08-25025 August 2014 Regarding Technical Specification Violation of Containment Integrity 05000285/LER-2014-004, Regarding Unqualified Limit Switches Render Safety Equipment Inoperable2014-06-20020 June 2014 Regarding Unqualified Limit Switches Render Safety Equipment Inoperable 05000285/LER-2014-003, Regarding Reactor Trip Due to Stator Water Cooling Leak During Maintenance2014-05-14014 May 2014 Regarding Reactor Trip Due to Stator Water Cooling Leak During Maintenance 05000285/LER-2013-014-01, 1 for Fort Calhoun Station Regarding Unqualified Components Used in Safety System Control Circuit2014-05-0202 May 2014 1 for Fort Calhoun Station Regarding Unqualified Components Used in Safety System Control Circuit 05000285/LER-2013-019-01, 1 for Fort Calhoun Regarding Non-Seismic Circulating Water Pipe Could Disable Raw Water Pumps2014-04-14014 April 2014 1 for Fort Calhoun Regarding Non-Seismic Circulating Water Pipe Could Disable Raw Water Pumps 05000285/LER-2014-002, Regarding Reactor Manual Trip Due to Control Rod Misalignment2014-03-12012 March 2014 Regarding Reactor Manual Trip Due to Control Rod Misalignment 05000285/LER-2014-001, Regarding Reactor Shutdown Due to Sluice Gate Failure2014-03-0707 March 2014 Regarding Reactor Shutdown Due to Sluice Gate Failure 05000285/LER-2013-015-01, Regarding Unqualified Coating Used as a Water Tight Barrier in Rooms 81 and 822014-02-14014 February 2014 Regarding Unqualified Coating Used as a Water Tight Barrier in Rooms 81 and 82 05000285/LER-2013-019, Regarding Non-Seismic Circulating Water Pipe Could Disable Raw Water Pumps2014-01-31031 January 2014 Regarding Non-Seismic Circulating Water Pipe Could Disable Raw Water Pumps 05000285/LER-2013-016, Regarding Reporting of Additional High Energy Line Break Concerns2014-01-0606 January 2014 Regarding Reporting of Additional High Energy Line Break Concerns 05000285/LER-2013-017, Regarding Containment Spray Pump Design Documents Do Not Support Operation in Runout2013-12-27027 December 2013 Regarding Containment Spray Pump Design Documents Do Not Support Operation in Runout 05000285/LER-2013-018, Regarding Postulated Fire Event Could Result in Shorts Impacting Safe Shutdown2013-12-26026 December 2013 Regarding Postulated Fire Event Could Result in Shorts Impacting Safe Shutdown 05000285/LER-2013-008-01, Regarding Previously Installed GE Iva Relays Failed Seismic Testing2013-12-18018 December 2013 Regarding Previously Installed GE Iva Relays Failed Seismic Testing 05000285/LER-2013-014, Regarding Unqualified Components Used in Safety System Control Circuit2013-12-18018 December 2013 Regarding Unqualified Components Used in Safety System Control Circuit 05000285/LER-2012-017-02, Regarding Containment Valve Actuators Design Temperature Ratings Below Those Required for Design Basis Accidents2013-12-0606 December 2013 Regarding Containment Valve Actuators Design Temperature Ratings Below Those Required for Design Basis Accidents 05000285/LER-2013-013, Unqualified Components Used in Safety System Control Circuit2013-12-0202 December 2013 Unqualified Components Used in Safety System Control Circuit 05000285/LER-2013-015, Regarding Unqualified Coating Used as a Water Tight Barrier in Rooms 81 and 822013-11-12012 November 2013 Regarding Unqualified Coating Used as a Water Tight Barrier in Rooms 81 and 82 05000285/LER-2012-009-01, Inoperable Equipment Due to Lack of Environmental Qualifications2013-10-31031 October 2013 Inoperable Equipment Due to Lack of Environmental Qualifications 05000285/LER-2013-010-01, Regarding HPSI Pump Flow Imbalance2013-10-23023 October 2013 Regarding HPSI Pump Flow Imbalance 05000285/LER-2012-015-01, Fort Calhouns Station Re Electrical Equipment Impacted by High Energy Line Break Outside Containment2013-09-30030 September 2013 Fort Calhouns Station Re Electrical Equipment Impacted by High Energy Line Break Outside Containment 05000285/LER-2013-012, Regarding Intake Structure Crane Seismic Qualification2013-09-30030 September 2013 Regarding Intake Structure Crane Seismic Qualification 05000285/LER-2013-006-01, Regarding Use of Teflon in LPSI and CS Pump Mechanical Seal2013-08-26026 August 2013 Regarding Use of Teflon in LPSI and CS Pump Mechanical Seal 05000285/LER-2013-007-01, Regarding Containment Air Cooling Unit (VA-16A/B) Seismic Criteria2013-08-16016 August 2013 Regarding Containment Air Cooling Unit (VA-16A/B) Seismic Criteria 05000285/LER-2013-011, The Fort Calhoun Station, Inadequate Design for High Energy Line Break in Rooms 13 and 19 of the Auxiliary Building2013-08-12012 August 2013 The Fort Calhoun Station, Inadequate Design for High Energy Line Break in Rooms 13 and 19 of the Auxiliary Building 05000285/LER-2013-005-01, Regarding Control Room HVAC Modification Not Properly Evaluated2013-07-31031 July 2013 Regarding Control Room HVAC Modification Not Properly Evaluated 05000285/LER-2013-010, Regarding HPSI Pump Flow Imbalance2013-07-0202 July 2013 Regarding HPSI Pump Flow Imbalance 05000285/LER-2012-016-01, Regarding Unanalyzed Charging System Socket Welds to the Reactor Coolant System2013-06-25025 June 2013 Regarding Unanalyzed Charging System Socket Welds to the Reactor Coolant System 05000285/LER-2013-009, Regarding Tornado Missile Vulnerabilities2013-06-14014 June 2013 Regarding Tornado Missile Vulnerabilities 05000285/LER-2013-008, The Fort Calhoun Station, Previously Installed GE Iav Relays Failed Seismic Testing2013-06-0707 June 2013 The Fort Calhoun Station, Previously Installed GE Iav Relays Failed Seismic Testing 05000285/LER-2013-007, Regarding Containment Air Cooling Units (VA-16A/B) Seismic Criteria2013-06-0404 June 2013 Regarding Containment Air Cooling Units (VA-16A/B) Seismic Criteria 05000285/LER-2012-019-01, Regarding Traveling Screen Sluice Gates Found with Dual Indication2013-05-14014 May 2013 Regarding Traveling Screen Sluice Gates Found with Dual Indication 05000285/LER-2012-018-01, Regarding Containment Air Cooling Units Not Properly Tested During Cycle 262013-05-0808 May 2013 Regarding Containment Air Cooling Units Not Properly Tested During Cycle 26 05000285/LER-2013-006, Use of Teflon in LPSI and CS Pump Mechanical Seals2013-05-0303 May 2013 Use of Teflon in LPSI and CS Pump Mechanical Seals 05000285/LER-2013-005, Regarding Control Room HVAC Modification Not Properly Evaluated2013-04-29029 April 2013 Regarding Control Room HVAC Modification Not Properly Evaluated 05000285/LER-2013-004, Regarding Inverters Inoperable During Emergency Diesel Generator Operation2013-04-23023 April 2013 Regarding Inverters Inoperable During Emergency Diesel Generator Operation 05000285/LER-2013-003, Regarding Calculations Indicate the HPSI Pumps Will Operate in Run-out During a DBA2013-04-0101 April 2013 Regarding Calculations Indicate the HPSI Pumps Will Operate in Run-out During a DBA 05000285/LER-2013-002, Regarding CVCS Class 1 & 2 Charging Supports Are Unanalyzed2013-03-26026 March 2013 Regarding CVCS Class 1 & 2 Charging Supports Are Unanalyzed 2017-05-11
[Table view] |
LER-2003-001, Re Lack of Guidance Results in Noncompliance with Technical Specification Surveillance Requirement |
Event date: |
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Report date: |
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Reporting criterion: |
10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications |
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2852003001R00 - NRC Website |
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wrsH Omaha Public Power Distnct 444 Southt 16thi Street Mall Omaha NE 68102-2247 May 5, 2003 LIC-03-0059 U. S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555-0001
Reference:
Subject:
Docket No. 50-285 Licensee Event Report 2003-001 Revision 0 for the Fort Calhoun Station Please find attached Licensee Event Report 2003-001, Revision 0, dated May 5, 2003. This report is'being submitted pursuant to 10 CFR 50.73(a)(2)(i)(B).
Attachment c:
E. W. Merschoff, NRC Regional Administrator, Region IV A. B. Wang, NRC Project Manager J. G. Kramer, NRC Senior Resident Inspector INPO Records Center Winston and Strawn Employment wvithl Equal Opportunity I
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Abstract
During the week of March 3, 2003 an evaluation to determine the adequacy of the Fort Calhoun Station (FCS) boric acid program was conducted. As part of the evaluation, one of the evaluation team members requested information on the results of the VT-2 ispections on the lower portion of the reactor vessel. A review of the inspection results indicated that the VT-2 examination had not been accomplished which is not in compliance with Section Xl of the ASME Boiler and Pressure Vessel Code as required by Technical Specification section 3.3.1.a.
The most probable cause of this event is lack of procedural guidance, caused by poor human factors in the FCS procedure that is used to inspect the rest of the reactor coolant system. There appears to have been a mind set among individuals that the room housing the reactor vessel was then, and had always been, inaccessible because of radiological dose considerations. Further, there appears to be a mind set that "inaccessible" because of radiological dose considerations is equivalent to "inaccessible" as defined in the code.
FCS will perform the examination as required by the code if the relief request submitted for processing prior to the 2003 refueling outage is delayed or denied.
NRC FORN 366 (7-2001)
(If more space is required, use additional copies of (If more space is required, use additional copies of NRC Forn 366A)
Reactor o
Head to Vessel Flange o
ICI Grayloc Flanges o
CEDM Tool Access Flanges o
CEDM Seal Housing Flanges" OP-ST-RC-3007 step 7.7.3 states, "Perform VT-2 visual inspection on all Class I piping associated with the items listed in Attachment 3". Attachment 3 lists numerous specific.valves associated with the RCS or associated systems. Nowhere in OP-ST-RC-3007 is there a specific requirement to perform a VT-2 visual inspection on the reactor vessel after a four hour "soaking" period. There are specific instructions to perform VT-2 inspections of unique locations of the RCS.
Earlier versions of the OP-ST-RC-3007 were reviewed. Revision 0 has an equivalent step to step 7.7.2 of revision 12 that states, "Inspect all RCS Class I components and piping including, but not limited to, those associated with the following:
Reactor Coolant Pumps (all)
Steam Generators (both)
Pressurizer (including heaters)
Reactor" In May, 1990, there was an effort to ensure that Inconel 600 penetrations were emphasized during the VT-2 inspections. Therefore, under each piece of major equipment listed in the step reflecting the four hour hold, subsections were added indicating specific portions of the major equipment in an attempt to improve OP-ST-RC-3007. Revision 2 was issued in an attempt to further clarify and standardize OP-ST-RC-3007. The step was changed, and the word "component" was apparently inadvertently left out of the step. The step only required the inspection of piping.
On March 7, 2003, a review of the reportability of this event was completed. This is being reported pursuant to 10 CFR 50.73(a)(2)(i)(B).
SAFETY SIGNIFICANCE
This issue has a negligible impact on the health and safety of the public as discussed below:
There is little possibility of an unacceptable flaw or flaw propagation in the region of the reactor vessel not examined by the surveillance test. A visual examination would only have detected a through-wall flaw, and would have provided no additional assurance for less than through-wall flaws of any size. This conclusion is supported by the following:
Possibility of Undiscovered Flaw The presence of an undiscovered flaw resulting in leakage from the reactor vessel is not probable. The portion of the reactor shell in question does not contain any materials other than the reactor shell material (i.e., there are no mechanical penetrations in the bottom of the FCS reactor pressure vessel). That is, there are no materials present that are susceptible to rapid crack initiation and propagation mechanisms under the reactor operating conditions, such as the Ni-Cr-Fe alloy 600 family of materials. Consequently, there is considered a low likelihood of an active cracking mechanism that would produce flaw initiation and growth.
(If more space is required, use additional copies of (If more space is required, use additional copies of NRC Form 366A) 1992 Reactor Vessel Inservice Inspection Per ASME section Xl requirements, the reactor vessel weld inspection was performed during the 1992 refueling outage. This inspection consisted of 176 automated ultrasonic exams and an interior visual exam of the reactor vessel welds and heat affected zones. The inspection found no rejectable indications in any of the reactor pressure vessel welds or heat affected zones. Twenty minor indications were found that were initially categorized as code recordable.
twelve (12) of these indications-were small-laminar-indications.that did not obscure the-backwall UT signal and could be immediately dispositioned as code acceptable.
two (2) were determined to have code acceptable signal level when re-exam was performed.
three (3) were found to have been caused by transducer lift-off due to a local surface irregularity.
three (3) were sized and found to be well within the code allowable acceptance criteria for weld indications.
Based on these results the vessel was found to be sound and satisfactory to be returned to service.
RCS Leak Rate Trending The FCS staff has assembled 9 cycles of unknown leakage data that demonstrate RCS operating threshold for boundary integrity. This trending of unknown leakage based on a threshold definition and category zones has been verified by previous cycles and validated by known leakage issues. These two categories are grouped into either connection (range from 0.075 to 0.2 gpm) or boundary integrity leakage (greater than 0.2 gpm) in facilitating an assessment of system performance. In monitoring of cycles 19, 20 and 21 (current) it is evident the boundary leakage threshold has not been challenged in the previous and/or current cycle. During the current cycle, the RCS leak rate has been below the threshold baseline.
Therefore, this issue has a negligible impact on the health and safety of the public.
CONCLUSION The most probable cause of this event is lack of procedural guidance, caused by poor human factors in procedure OP-ST-RC-3007. There appears to have been a mind set among many individuals that the room housing the reactor vessel was inaccessible. When questioned why they thought the room was inaccessible, most individuals stated that it-was because of-radiological dose reasons.-Further,-there appears.to be a-mind -set that--
"inaccessible" because of radiological dose considerations is, in some way, equivalent to "inaccessible" as defined in the code. Some individuals assumed there was an engineering argument that documented the inaccessibility of the room as related to the code. There was an additional argument/mind set from some individuals that, since the word "component" was not included in the test, the reactor vessel inspection was not required. Therefore, it appears that a contributing cause for this event was a number of inappropriate mind sets.
CORRECTIVE ACTIONS
FCS will perform the examination as required by the code if the relief request submitted for processing prior to the 2003 refueling outage is delayed or denied. Other corrective actions related to this issue are being completed in accordance with the FCS corrective action program.
SAFETY SYSTEM FUNCTIONAL FAILURE This event did not result in a safety system functional failure in accordance with NEI 99-02.
PREVIOUS SIMILAR EVENTS
FCS has not had any previous similar events.
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