05000302/LER-1996-001, :on 960110,personnel Error Results in Operation Outside 10CFR50,App R Design Basis.Brought Affected Circuits Back Into Compliance W/App R & Completed Field Validation of Selected App R Drawings
| ML20148Q547 | |
| Person / Time | |
|---|---|
| Site: | Crystal River  |
| Issue date: | 06/23/1997 |
| From: | Catchpole T, Holden J FLORIDA POWER CORP. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| 3F0697-16, 3F697-16, LER-96-001, LER-96-1, NUDOCS 9707070134 | |
| Download: ML20148Q547 (7) | |
| Event date: | |
|---|---|
| Report date: | |
| Reporting criterion: | 10 CFR 50.73(a)(2)(ii)(B), Unanalyzed Condition |
| 3021996001R00 - NRC Website | |
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Fl:rida E939Y Cryste meer vidt 3 Docket No.60-302 i
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June 23,1997 3F0697-16 U. S. Nuclear Regulatory Commission Attention:
Document Control Desk Washington, D. C.
20555-0001
Subject:
Licensee Event Report (LER) 96-001-02 l
Dear Sir:
1 Please find the enclosed Licensee Event Report (LER) 96-001-02.
This supplemental report is submitted by Florida Power Corporation in accordance with i
10 CFR 50.73. It provides an update of corrective actions relative to the field validation of Appendix R drawings and a self assessment of configuration controls.
Sincerely,
. J. Holden, Director Nuclear Engineering and Projects JJH/TWC Attachment xt: Regional Administrator, Region 11 Project Manager, NRR Senior Resident Inspector e
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CRYSTAL RIVER UNIT 3 (CR-3) 0l5l0l010l310l2 1 lOFl 0 l 6 i
-~firLt (ai Personnel Error by Contractor Results in Operation Outside 10 CFR 50 Appendix R Design Basis EVENT DATE (5)
LER NUMBER (6)
REPORT DATE (7)
OTHER FACILITIES INVOLVED (R)
SEQUENTIAL REVISION FACILITY NAMES DOCKET NUM8ER(S)
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N'ME TELEPHONE NUMBER APEA CODE T.W. Catchpole, Sr. Nuclear Licensing Engineer 3l 5 l 2 5l 6 l 3 l-l 4l 6 l 0 l 1 COMPLETE ONE LINE FOR EACH COMPONENT FAILURE IN THIS REPOHT (13)
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On January 10, 1996, Florida Power Corporation's (FPC) Crystal River Unit 3 (CR-
- 3) was in MODE THREE (HOT STANDBY). During a walkdown to develop a modification for upgrading CR-3's Thermo-Lag fire barriers, an Appendix R separation criteria deviation was identified in that two conduits containing circuits for controlling "B" Train Emergency Feedwater flow to the steam generators were noted to pass through the same fire area as "A" Train circuits with no fire barrier protection in that area.
Although the condition was initially determined not reportable, upon further review on January 11, 1996 while the unit was in MODE FOUR (HOT SHUTDOWN), a 4-hour prompt notification was made to identify a degraded condition while the plant was shutdown. This report documents the separation problem as a condition outside CR-3's design basis. A justification for continued operation was established based on the existence of continuous roving fire watches in the affected area.
The cause of this event was cognitive personnel error by personnel involved in preparing a modification to install conduit and cable during a 1985 refueling outage.
Affected circuits have been brought back into compliance with Appendix R and a field validation of selected Appendix R drawings l
has been completed. A self-assessment of configuration controls covering multi-I discipline modifications was al so performed resulting in strengthening of l
procedural controls and review board oversight.
NFr Form 366(6-89)
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PAGE (3) bEQUENTIAL RIVl8KJN CRYSTAL RIVER UNIT 3 (CR-3)
YEAR NUMBER NUMBER ol sl ol ol ol 3l ol 2 ele olol1 0l2 o l 2 lOFl o l 6 rExr w n. -. - ecr uu.on EVENT DESCRIPTJ_0J On January 10, 1996, Florida Power Corporation's (FPC) Crystal River Unit 3 (CR-
- 3) was in MODE THREE (HOT STANDBY). During a walkdown by personnel involved in l developing a modification for upgrading CR-3's Thermo-Lag fire barriers, it was noted that conduits EFS56 and EFS57, which contain "B"
Train circuits for Emergency Feedwater System [BA] (EFW) components, were routed through the "A" 480-Volt Engineered Safeguards (ES) Switchgear Room.
The portion of the "B" Train conduits in the "A" Switchgear Room were not protected by fire barriers and l are within 20 feet of redundant "A" train circuits. See Figure 1.
Conduits EFS56 and EFS57 contain circuits which provide flow control signals and motive power to controllers [BA,FC] for the "B" train Emergency Feedwater Flow Control Valves [BA,FCV).
The redundant "A" train circuits are associated with controllers for "A" train Emergency Feedwater Flow Control Valves.
A Problem Report was generated to identify a noncompliance with 10CFR50 Appendix R, Section III.G.2.b separation criteria. The Problem Report was initially considered not l to be reportable on the basis that the conduit was in the confirmed route of the continuous roving fire watch established in response to NRC Bulletin 92-01 "Fatlure of Thermo-lag 330 Fire Barrier System to Perform its Specified Fire Endurance Function".
Upon further review of the Problem Report on January 11, 1996, while the unit was in MODE FOUR (HOT SHUTDOWN) it was determined the condition warranted a 4-hour report in accordance with 10CFR50.72. The installed condition did not conform with 10CFR50 Appendix R during the time period from 1985, when the conduits were installed, to the establishment of the compensatory fire watch in 1992. The notification was made at 1828 hours0.0212 days <br />0.508 hours <br />0.00302 weeks <br />6.95554e-4 months <br /> in accordance with 10CFR50.72(b)(2)(i) as a degraded condition found while the reactor was shutdown and was assigned Event Number 29826.
This report is being submitted in accordance with 10CFR50.73(a)(2)(ii)(B) to describe a condition outside the design basis of CR-3 with regard to 10CFR50 Appendix R Fire analyses during the time period described above.
EVENT EVAL.UATION l
The control circuits for the EFW Flow Control Valves are necessary for safe shutdown and are subject to the requirements of 10CFR50 Appendix R Section III.G.
This section addresses requirements for protection features needed to ensure one train of a safe shutdown system is free of fire damage.
CR-3's design basis requires separation by a fire barrier with a 3-hour rating, or separation by horizontal distance of more than 20 feet with no intervening combustibles or fire hazards, or enclosure in a fire barrier having a 1-hour rating.
In the second j
and third cases, the area must also be equipped with fire detection and automatic l
fire suppression. CR-3 complied with the enclosure requirements by use of a fire j
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PAGE (3) seautN % p navissoN CRYSTAL RIVER UNIT 3 (CR-3)
YEAR NWEER NWBER ol 5l ol ol ol 3l ol 2 ole ofol1 0l2 o l 3 lOFl 0 l 6 TEXT y,.
o..,,No.,.-,m resistive material called Thermo-Lag 330. AsshownonFigure1,ConduitsEFS56l and EFS57, which contain control power and instrumentation circuits for the "B" train EFW flow control valves, were routed with no protection through the "A" 480 l
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Volt Engineered Safeguards (ES) Switchgear Room within 20 horizontal feet of the controllers for the "A" train control valves. The 480V Switchgear Rooms are 3-hour fire areas. "cnduits EFS56 and EFS57 were intended to be routed to the "B" 480 Volt ES Switcngear Room through the corridor area from EFIC Room "B".
The Corridor area is a 1-hour fire area and the portion of EFS56 and EFS57 in this area is coated with Thermo-Lag. If there was a fire in the "A" 480 ES Switchgear j
Room in the area of the EFV-57 and EFV-58 controllers and the as-installed conduits EFS56 and EFS57, then flow control or level fill rate of emergency feedwater could'be lost.
An operability assessment of the EFW System was conducted in accordance with Compliance Procedure CP-150, " Identifying and Processing Operability Concerns."
The assessment provided a jcstification for continued operation which recognized l I
the existence of continuous reving fire watches that include the "A" 480 Volt ES Switchgear Room.
These fire oatches had been established in response to NRC concerns which showed Thermo-Lig material to be deficient.
The operability assessment also noted the fire watch in the "A"
480V ES Switchgear Room supplements the installad Pyotronics fire detection system in the area.
The Shift Supervisor on Duty (S500) iccepted the operability assessment on January 15, 1996 and closed the concern.
CAUSE
The primary cause of this event was cognitive personnel error by engineering l personnel.
Conduits EFS56 and EFS57 were installed in 1985 during Refuel 5 as part of a field change to modification 80-10-66-08A "EFIC Electrical Conduit and l Cable Installation".
The field change was a multi-discipline (electrical and structural) change initiated by the electrical discipline for the purpose of addressing Appendix R.
After the electrical design was completed, structural engineering completed the balance of the design for the conduit installation and approved the change notice. The structural design engineer selected a different route for conduit supports from that depicted on layout sketches provided by electrical engineering.
It is surmised that this route was selected to take advantage of existing conduit supports and/or due to interferences present in the route selected by electrical engineering. The structural engineer was apparently not aware of the impact this change had on Appendix R criteria. Further, design verification of the field change did not include a final review by electrical engineering.
Two other factors may have contributed to this event or the failure to discover l the discrepancy.
One was insufficient procedures in place during Refuel 5 to guide interface between engineering disciplines.
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PAGE (3)
CRYSTAL RIVER UNIT 3 (CR-3)
YEAR ue 0l6l0l0l0l3l0l2 9l6 0l0l1 0l2 0 l 4 lOFl 0 l 6 TEXT Nt awe spece a reeuwed vee ad Dwane NAC Fwe 35dA e (1I) a multi-discipline review and co-approval signatures of the Design Engineers and Verification Engineers involved in developing the modification. Another factor was that Appendix R drawings were developed using conduit layout drawings which are diagrammatic and do not depict the exact route.
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IMMEDIATE CORRECTIVE ACTION
When the Problem Report for this condition was first presented to the Shift Manager on January 10, 1996 it was confirmed the separation discrepancy was in a fire area covered by the roving fire watch.
ADDITIONAL CORRECTIVE ACTION 1.
Drawings which indicate the routing of Conduits EFS56 and EFS57 have been revised to depict the correct routing.
l 2.
FPC conducted a validation of Appendix R drawings using a sample-based l approach to compare the as-built configuration against the configuration confirmed by field walkdown. The sample involved a horizontal and vertical validation of all fire barrier protected raceways associated with the Emergency Feedwater system and Fire Areas CC-108-107 and CC-124-116 which are the 4160V 'B' and 480V 'B' Switchgear Rooms, respectively.
Although additional discrepancies were identified consisting of tray labeling and color coding problems, small breaks in firewrap, and information missing from drawings, none were representative of design basis issues similar to those described in this report.
Based on the results of the validations completed, additional walkdowns were determined to be unnecessary.
3.
Circuits contained in Conduit EFS56 were relocated and wrapped with Mecatiss l per the commitment described in FPC letter 3F1295-05 to NRC dated December 21, 1995.
Circuits contained in Conduit EFS57 were rerouted such that firewrap is no longer required.
ACTION TO PREVENT RECURRENCE Bothshort-termandlong-termactionshavebeenidentifiedtoaddresstherootl cause for this event. A copy of the initiating Problem Report has been provided to Nuclear Engineering Design personnel to re-emphasize the need for proper design interface between engineering disciplines. In addition, a self-assessment was performed of the effectiveness of existing (enhanced) configuration controls in the area of design interface between engineering disciplines. The results of the self assessment indicated that the existing interface controls needed to be enhanced.
FPC has taken the following actions to enhance interdisciplinary design controls:
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DEQVENT A REVIS4CN CRYSTAL RIVER UNIT 3 (CR-3)
YEAR NWHR NWHR 0l 5l 0l 0l 0l 3l 0l 2 9l6 0l0l1 0l2 0 l 5 lOFl 0 l 6 TEXT (a mo<e o a m <egwed use **m=av Nac ra<m med * (17) 1.
Utilization of the precursor card process has been increased for documenting and resolving configuration and design issues.
2.
A Design Review Board (DRB) has been established consisting of interdisciplinary members from the design engineering, systems engineering, operations, maintenance, training, and licensing organizations. Generally, projects which have an effect on nuclear safety will be selected for DRB review.
3.
Nuclear Engineering Procedure (NEP) 104, " Interface Design Control," has been revised to strengthen expectations regarding the need for verification of the final design by the submitting design discipline when inputs are provided to other disciplines.
PREVIOUS SIMILAR EVENTS
LER 89-39 reported a similar event. In this LER, it was discovered that a fuse was relocated to satisfy Appendix R requirements; however, the new location did not fulfill the desired objective.
The cause of the separation problem was identified as design error in that personnel preparing and reviewing a field change for installation in Refuel 5 were not knowledgeable of Appendix R Separation criteria. The actions to prevent recurrence as identified in the LER were to review all modifications subsequent to Refuel 5 to assure Appendix R separation criteria was met.
The evaluation of the problem report which initiated the LER noted it was considered to an isolated case of misinterpretation of Appendix R design requirements by field personnel.
1 ATTACHMENT I
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