05000302/LER-1997-001, :on 970128,cavitation from Inadequate NPSH Affected Availibility of Efp.Cause by Ineffective Configuration Change Management.Edg a Power Upgraded,Efw Sys Mods & Failure Modes & Effects Analyzed
| ML20135C743 | |
| Person / Time | |
|---|---|
| Site: | Crystal River  |
| Issue date: | 02/27/1997 |
| From: | Catchpole T, Holden J FLORIDA POWER CORP. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| 3F0297-26, 3F297-26, LER-97-001, LER-97-1, NUDOCS 9703040258 | |
| Download: ML20135C743 (9) | |
| Event date: | |
|---|---|
| Report date: | |
| Reporting criterion: | 10 CFR 50.73(a)(2)(ii)(B), Unanalyzed Condition 10 CFR 50.73(a)(2)(vii), Common Cause Inoperability 10 CFR 50.73(a)(2)(v), Loss of Safety Function |
| 3021997001R00 - NRC Website | |
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Florida Pswer Ew"unn a oocim uo. soma February 27,1997 3F0297-26 U. S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, D. C. 20555-0001
Subject:
Licensee Event Report (LER) 97-001-00
Dear Sir:
Please find the enclosed Licensee Event Report 97 001-00 which discusses an unanalyzed f
condition which could have rendered the Emergency Feedwater System incapable of fulfilling its intended safety and accident mitigation functions.
This report is submitted by Florida Power Corporation in accordance with 10 CFR 50.73.
Sincerely, J. J. Holden, Director Nuclear Engineering and Projects JJH/TWC Attachment xc:
Regional Administrator, Region Il Project Manager, NRR p,7 Senior Resident inspector
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PAGE (3)
CRYSTAL RIVER UNIT 3 (CR-3) o l 5 l 0 l 0 l 0 l 310 l 2 1 loFl 0 l 8 TITLE (4)
Ineffective Change Managernent Results in Unrec0gnized NPSH lssue Affecting Emergency Feedwater Availability EVENT DATE (6)
LER NUMBER (6)
HEPORT DATE (7)
OTHER FACILITIES INVOLVED (8)
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LICENSE CONTACT FOR THul LER(12)
NAME TELEPHONE NUMBER AREA 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 REPORT (13)
CC.USE SYSTEM COMPONENT MANUFAC-REPORTABLE
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On January 28,1997, Florida Power Corporation's (FPC) Crystal River Unit 3 (CR-3) was in MODE 5 (COLD SHUTDOWN). Discussions with NRC inspection personnel identified that FPC had not explicitly reported a condition that existed prior to May,1996 involving inadequate net positive suction head (NPSH) affecting one of the two Emergency Feedwater Pumps (EFP). On a loss of 'B' DC power, the turbine-driven pump's (EFP-2) flow control valves would remain fully open and EFP-2 would start in a maximum flow condition resulting in cavitation from inadequate NPSH which could lead to pump f ailure. In addition to an Emergency Diesel Generator load management concern, the postulated loss of 'B' DC power single failure coincident with a Small Break Loss of Coolant Accident and Loss of Offsite Power, a low probability event, could have resulted in two situations in which emergency feedwater may not have been available to perform its intended safety and accident mitigation functions. These include a design feature which trips the motor driven pump, EFP-1 at a Reactor Coolant System pressure of 500 pounds per square inch gauge, and a point in time at which EFP-1 would need to be secured in order to load the Low Pressure Injection pump onto the EDG in order to provide adequate NPSH to the High Pressure injection pump. As a result, CR-3 was in an unanalyzed condition which could have rendered the emergency feedwater system incapable of fulfilling its intended safety and accident mitigation functions.
The cause of this event was ineffective configuration change management. Corrective Actions include a power upgrade of the 'A' EDG, EFW system modifications to eliminate NPSH concerns, and a failure modes and effects analysis.
NRC Form 386 (6-89)
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EVENT DESCRIPTION
On January 28,1997, Florida Power Corporation's (FPC) Crystal River Unit 3 (CR-3) was in MODE 5 (COLD SHUTDOWN). The unit has been in shutdown since September,1996. FPC management decided to voluntarily keep the plant shutdown until concerns with various design related issues were resolved. Discussions with NRC inspection personnel identified that FPC had not explicitly reported a condition that existed prior to May,1996 involving inadequate net positive suction head (NPSH) affecting one of the two Emergency Feedwater Pumps [BA,P](EFP). This condition was described as an initiating event in two recently submitted event reports, LER 96-020-00 regarding Emergency Diesel Generator [EK,DGl(EDG) loading issues and LER 96-024-01 which discussed an unanalyzed condition regarding Emergency Feedwater (EFW) availability.
See Previous Similar Events for additional information.
The steam-driven emergency feedwater pump, EFP-2, would not be able to perform its intended safety function after a postulated failure of the 'B' DC bus coincident with a Loss of Offsite Power (LOOP). The 'B' EDG would not start due to reliance on the 'B' DC system [EJ,BTRY). No AC or DC puer would be available to 'B' train Engineered Safeguards (ES) components. However, the 'A' Emergency Feodwater initiation and Control [JB](EFIC) train would open one of the two redundant steam admission valves, Auxiliary Steam Valve [SA,lSV) ASV-204, which provides motive steam for EFP-2. Due to the loss of 'B' DC power, the EFP-2 flow control valves [BA,FCV) would remain fully open and EFP-2 would start in a maximum flow condition resulting in cavitation from inadequate NPSH which could lead to pump failure. This NPSH concern was initially determined to affect only one train of EFW.
During the root cause analysis and other investigations in support of LER 96-024-01, near the end of 1996, it became evident that a dependency existed in which EFP-2 was relied upon to support EDG-1 A operability. The motor-driven pump, EFP-1, would also be unavailable either when the Low Pressure injection [BP](LPI) actuation occurs or when the High Pressure injection /LPI " piggyback" mode would need to be established. These situations are exp ained further in the Event Evaluation section. Therefore, in certain scenarios, a single failure (loss of 'B' DC power) rendering EFP-2 unavailable could result in both trains of EFW and possibly the 'A' EDG inoperable / unavailable.
The above condition existed from December 1987 when ASV-204 was powered and received its open signal from the 'A' EFIC system until this EFIC signal was removed in May,1996 during Refueling Outage 10. As a result, CR-3 was in an unanalyzed condition which could render EFW incapable of fulfilling its intended safety and accident mitigation functions. This unanalyzed condition is reportable under 10CFR50.73 (a)(2)(ii)(B) as a condition outside design basis and 10CFR50.73(a)(2)(vii)(D) as an event where a single condition caused two independent trains to become inoperable in a single system designed to mitigate the consequences of an accident.
NRC Fem 366A (6-89)
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- - SEQUENTIAL REV18 EON YEAR NWBER
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EVENT DESCRIPTION
On January 28,1997, Florida Power Corporation's (FPC) Crystal River Unit 3 (CR-3) was in MODE 5 (COLD SHUTDOWN). The unit has been in shutdown since September,1996. FPC managon ent decided to voluntarily keep the plant shutdown until concerns with various design ta ated issues were resolved. Discussions with NRC inspection personnel identified that FPC had not explicitly reported a condition that existed prior to May,1996 involving inadequae net positive suction head (NPSH) affecting one of the two Emergency Feedwater Pumps [BA,P](EFP). This condition was described as ari initiating event in two recently submitted event reports, LER 96-020-00 regarding Emergency Diesel Generator [EK,DGl(EDG) loading issues and LER 96-024-01 which discussed an unanalyzed condition regarding Emergency Feedwater (EFW) availability.
See Previous Similar Events for additional information.
The stec. -driven emergency feedwater pump, EFP-2, would not be able to perform its intended safety function after a postulated failure of the 'B' DC bus coincident with a Loss of Offsite Power (LOOP). The 'B' EDG would not start due to reliance on the 'B' DC system [EJ,BTRY]. No AC or DC power would be available to 'B' trein Engineered Safeguards (ES) components. However, the 'A' Emergency Feedwater Initiation and Control [JB](EFIC) train would open one of the two redundant steam admission valves, Auxiliary Steam Valve [SA,lSV) ASV-204, which provides motive steam for EFP-2. Due to the loss of 'B' DC power, the EFP-2 flow control valves IBA,FCV) would remain fully open and EFP-2 would start in a maximum flow condition resulting in cavitation from inadequate NPSH which could lead to pump failure. This NPSH concern was initially determined to affect only one train of EFW.
During the root cause analysis and other investigations in support of LER 96-024-01, near the end of 1996, it became evident that a dependency existed in which EFP-2 was relied upon to support EDG-1 A operability. The motor-driven pump, ?FP-1, would also be unavailable either when the Low Pressure Injection [BP](LPI) actuation occurs or when the High Pressure injection /LPI " piggyback" mode would need to be established. These situations are explained further in the Event Evaluation section. Therefore, in certain scenarios, a single failure (loss of 'B' DC power) rendering EFP-2 unavailable could result in both trains of EFW and possibly the 'A' EDG inoperable / unavailable.
The above condition existed from December 1987 when ASV-204 was powered and received its open signal from the 'A' EFIC system until this EFIC signal was removed in May,1996 during Refueling Outage 10. As a result, CR-3 was in an unanalyzed condition which could render EFW incapable of fulfilling its intended safety and accident mitigation functions. This unanalyzed condition is reportable under 10CFR50.73 (a)(21.!ii)!9) as a condition outside design basis and 10CFR50.73(a)(2)(vii)(D) as an event where a single condition caused two independent trains to become inoperable in a single system designed to mitigate the consequences of an accident.
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CRYSTAL RIVER UNIT 3 (CR-3)
. NUWeER YEAR NUMBER I
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BACKGROUND 1 SYSTEM DESCRIPTION I
The Emergency Feedwater (EFW) system provides secondary coolant to the Once Through Steam Generators [AB,SGl(OTSG) in the event the Main Feedwater System [SJ](FW) is rendered inoperable and is unable to perform this function.
Tne EFW system has two equipment trains (See Figure 1). Each train is capable of feeding both OTSG's. The two trains taka suction from a common line. The flow control valves associated with each pump operat; cn DC power. The valves are normally open and require DC power to close and are open in the standby mode. EFP-1 is motor-driven and is aligned to the 'A' Emergency Diesel Generator during LOOP conditions. EFP-2 is turbine-driven; motive steam is supplied from the Main Steam [ sal (MS) header. The system includes two valves, ASV-5 and ASV-204, which open to admit steam to EFP-2 when EFW actuates. The valves are installed in parallel with one another. Only one of the valves must open in order to start the pump. ASV-5 receives an OPEN command from an actuation of the 'B' EFIC train.
ASV-204 was installed in 1985, was powered from
'B' Class 1E power sources, and received its OPEN command from the 'B' EFIC train.
In December 1987, FPC moved the ASV-204 power supply and OPEN command to the 'A' side Class 1E sources so the valve opened on the
'A' EFIC actuation train signal, thus allowing use of EFP-2 for 'A' EDG load reduction. This configuration was established so that, during a Loss of 'B' DC bus, EFP-2 could be run in parallel with EFP-1, the motor-driven Emergency Feedwater pump.
With both pumps sharing secondary coolant flow to the OTSG's, the electrical load or the motor driven pump on the 'A' EDG was reduced. This modification was implemented to reduce EDG demand if there was a loss of 'B' train power.
Even with this load reduction achieved in December 1987, the 'A' EDG did not have the capacity to support both EFP-1 and the Low Pressure injection [BP,P](LPl) pump concurrently.
A modification was installed in June,1990 to trip EFP-1 and start the LPI pump when RCS pressure dropped below 500 pounds per square inch gauge (psig) when the EDG was powering the 'A' bus.
EVENT EVALUATION This event becomes a concern in certain scenarios when EFW is required. The EFW system was designed to handle several abnormal events including Loss of main feedwater, Loss of main feedwater with loss of offsite power, Loss of main feedwater with loss of offsite and on-site AC power, Main feedwater line break, Main steam line break /EFW line break, and Small Break Loss of Coolant Accident (SBLOCA).
NRC Fc,m 366A (6-49)
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. REVISION CRYSTAL RIVER UNIT 3 (CR-3)
YEAR NWBER NWBER 0l 5l 0l 0l 0l 3l 0l 2 9l7 0l0l1 0l0 0 l 4 loFl 0 l 8 TEXT gn neae space e reeamt uen eemenemnc Form seen e (17)
The scenario of concern in this event involves a single failure of the 'B' DC bus coincident with a SBLOCA and Loss of Offsite Power (LOOP). This scenario does not impact the current Probabilistic Safety Analysis (PSA) analyses for core damage due to the extremely low frequency of occurrence. The probability of a SBLOCA with a LOOP and a Loss of 'B' DC bus is 8.3x10E-11 per year.
Historically, for SBLOCA's, FPC's nuclear steam system supplier, Framatome Technologies, Inc. (FTI), formerly Babcock & Wilcox, maintained that mitigation of the transient with acceptable consequences could be demonstrated with only one train of Emergency Core Cooling System (ECCS) available: one High Pressure injection (HPI) pump [BQ,P], one LPI pump, and one EFW pump. However, with only one EDG available providing power to 'A' train components and failure of EFP-2, there are two situations in which the remaining Emergency Feedwater pump, EFP-1, would not be available.
One situation wherein EFW would not be available occurs at the point in which the Reactor Coolant System (RCS) depressurizes to 500 psig at wh:ch time EFP-1 is tripped due to the LPl/EFP-1 trip block modification discussed in the Background section.
Another situation could occur if the Borated Water Storage Tank [BP,TK](BWST) had to be isolated before RCS pressure was reduced below the maximum discharge pressure for the LPI pumps. During a design basis LOCA, the Reactor Building Spray [BE](BS), LPI, and HPI systems are automatically aligned to obtain suction from the BWST. As inventory is lost through the break, it accumulates in the Reactor Building [NH](RB) Sump. After the BWST is drained to the swapover level, ECCS pump suction is transferred to the RB Sump. If this situation occurred, it would be necessary to place the HPI-LPI systems into the " piggy-back" mode of operation. This is the mode in which LPI pumps take suction from the sump in order to provide adequate NPSH to the HPl pumps. In order to load the LPI pump onto the EDG, EFP-1 would have to be secured. With EFP-1 secured, no feedwater would be provided to the OTSG's.
Emergency Operating Procedures (EOP's) do not provide guidance to maintain emergency feedwater for this case. This operator action to place the LPI pump in the piggyback alignment is necessary to satisfy the long term core cooling requirements specified in 10 CFR 50.46.
In either of the above cases, FPC would be unable to ensure compliance with 10CFR50.46 acceptance criteria for Emergency Core Cooling Systems. In certain postulated scenarios, peak clad temperatures may exceed regulatory limits. A loss of heat transfer from the core will result in increasing fuel and cladding temperature which, if not mitigated, will result in fuel uncovery and damage.
In addition to the above, even if the RCS stays above 500 psig and the HPl/LPI " piggy-back" arrangement is not required, at some point after the event, operators must apply certain NRC Fo,m 366A (6-89)
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SEQUENTIAL REVISION CRYSTAL RIVER UNIT 3 (CR-3) 0l 5l 0l 0l 0l 3l 0l 2 9l7 0l0l1 0l0 0 l 5 joFl 0 l 8 TEXT tr mere speoe e reened use em*rerei nrnC Form 366A e (17) manual loads to the 'A' EDG to address Control Complex (CC) cooling concerns. The EDG loading calculations assume the Control Complex Emergency Duty Supply Fan [VI, FAN], the CC Return Air Fan, and the CC EFIC Room Fan are manually loaded at 30 minutes. The same calculations assume the Control Complex Chillers (NA,KM,CHUl are manually loaded at one hour. The Chilled Water System is used to maintain the Control Room and other enclosures within the control complex, particularly those which contain electronic components, at a temperature / humidity level that affords personnel comfort and is compatible for electronic equipment. With EFP-1 supplying the entire EFW load, the resulting 'A' EDG kilowatt (KW) loading could be increased by greater than 200 KW which would allow the 'A' EDG to remain within its design rating but not provide sufficient margin to allow all of the additional manual loads to be added. Analyses indicate there may be sufficient margin to accommodate the CC Emergency Duty Supply fans, but not the EFIC 'A' Room Fan and the CC chillers. Presuming operators would not be able to manage additional EDG toads, not having the EFIC 'A' Room Fan and the CC chillers would result in increased temperature beyond the qualified operating conditions of vital plant instrumentation.
CAUSE
The cause of this event was ineffective configuration change management. As noted in CR-1 3's Phase 11 Management Corrective Action Plan (MCAP ll), there was a heavy reliance upon Architect-Engineer, contractor, and NSSS resources for performance of design activities for the first eighteen years of plant operation. As a result, there was ineffective technology transfer from the external sources to CR-3 engineers. Specifically, reliance on EFP-2 and the effects of loss of DC power scenarios were not fully understood.
IMMEDIATE CORRECTIVE ACTION
Due to the EFW/EDG issues, and other design related issues, FPC management made a decision to voluntarily keep the plant shut down until these issues are adequately addressed.
FPC has developed MCAP 11 to communicate management expectations and provide direction in several areas of plant performance.
For reference purposes, the following additional corrective actions are identified as applicable with MCAP 11 Action Item designations.
In addition, FPC formed a Restart Panel patterned after the NRC Inspection Manual Chapter 0350 " Staff Guidelines for Restart Approval" process to manage actions necessary to safely return CR-3 to power operation and ensure subsequent reliable operation. The following additional corrective actions are identified as applicable, with Restart issue numbers.
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YEAR NUMBER NUMBER 0l5l0l0l0l3, 0l2 9l7 0l0l1 0l0 0 l 8 lOFl 0 l 8 l
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A power upgrade for the
'A' EDG will be accomplished and appropriate EFW system modifications such as installation of cavitating venturis will be implemented to eliminate i
NPSH concerns and reduce operator burden prior to restart from the current voluntary outage.
(FPC Restart issues D-5 and D-6).
A Failure Modes and Effects Analysis of the LOCA, LOOP and Loss of DC Power scenario has been initiated and will be completed prior to restart. (MCAP Action C-CC1-1).
ACTION TO PREVENT RECURRENCE A " stand down" was implemented in Nuclear Operations Engineering (NOE) to emphasize the j
importance of improving safety culture. (MCAP Action B-RC1-1).
)
1 Engineering staffing levels have been increased to attract talent from outside FPC that can i
l increase design competency, (MCAP Action B-RC1-7).
A directive has been issued to restore system design margins primarily through physical means (modification or testing) as opposed to analytical means. (MCAP Action B-RC1-8).
PREVIOUS SIMILAR EVENTS
There has been one previous event involving the EFW system reported in accordance with 10CFR50.73(a)(2)(v) in which the condition was determined to have prevented the fulfillment of a safety function. LER 85-027 reported a condition wherein the steam-driven EFP was disabled per procedure and the motor-driven pump was disabled due to a spurious EFIC actuation while calibrating EFIC instrumentation.
A second spurious actuation occurred i
resulting in no EFW response.
LER's94-006, 95-015, and 95-016 reported setpoints for EFIC system instrumentation determined to be non-conservative relative to revised analyses using new setpoint i
methodology which resulted in questioning the system's ability to perform its intended safety function.
l On October 10,1996, FPC provided a voluntary LER (96-020-00) to describe an unreviewed l
l safety question (USO) involving the EDG loading calculation that was developed in support of the plant modification which removed the automatic open signal from ASV-204.
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.a e we v.,. ua. (m On November 12,1996, FPC issued LER 96-024-00, subsequently supplemented on February 14, 1997, to report an unanalyzed condition regarding emergency feedwater unavailability below 500 psig RCS pressure created as a result of the plant modification implemented in May 1996 which removed the automatic open signal frorn ASV-204.
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