05000333/LER-1999-001, :on 990114,incorrect EDG line-up During Fire Placed Plant in Condition Outside Design Basis.Caused by Inadequate Procedures & Insufficient Operator Training. Revised EDG Operating & Surveillance Test Procedures
| ML20205G248 | |
| Person / Time | |
|---|---|
| Site: | FitzPatrick  |
| Issue date: | 03/31/1999 |
| From: | Steigerwald R POWER AUTHORITY OF THE STATE OF NEW YORK (NEW YORK |
| To: | |
| Shared Package | |
| ML20205G240 | List:
|
| References | |
| LER-99-001, LER-99-1, NUDOCS 9904070181 | |
| Download: ML20205G248 (19) | |
| Event date: | |
|---|---|
| Report date: | |
| Reporting criterion: | 10 CFR 50.73(a)(2)(ii) 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications |
| 3331999001R00 - NRC Website | |
text
1 NRC FORM 366 U.S. NUCLEAR REaVLATORY COMMISSION APPROVED BY OMB NO. 3150-0104 EXPIRES 06/30/2001 i,5-1998)
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an information coli ction does not display a currentiv valid i
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James A. FitzPatrick Nuclear Power Plant 05000333 1 OF 19 i
l TITLE 14)
In:orrect Emergency Diesel Generator Line-Up During Fire Placed Plant in Condition Outside Design Basis EVENT DATE (5)
LER NUMBER (6)
REPORT DATE (7)
OTHER FACILITIES INVOLVED (8) i F ACILITY NAME DOCKET NVMBER SEOU AL E S N MONTH DAY YEAR YEAR MONTH DAY YEAR p
FACfLITV NAME DOCKET NUMBER 01 14 99 99 001 01 03 31 99 N/A 05000 I
THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR 5: (Check one or more) (11)
OPERATING N
20.2201(b) 20.2203(a)(2)(v) x 50.73taH2Hi) 50.73(a)(2Hviii)
MODE (9)
POWER 20.2203(aHI) 20.2203(a)(3)(i) x 50.73(aH2Hii) 50.73(aH2Hx) 100 LEVEL (10) 20.2203(aH2Hi>
20.2203(a)(3Hii) 50.73(aH2Hiii) 73.71 20.2203(aH2Hii) 20.2203(aH4) 50.73(aH2)(iv) x OTHER 20.2203(aH2Hiii) 50.36(c)(1) 50.73(aH2Hv)
Specify in Abstract below 20.2203(aH2Hiv) 50.36(c)(2) 50.73(aH2)(vii) or in NRC Form 366A LICENSEE CONTACT FOR THIS LER (12)
NAME TELEPHONE NUMBtH (inciuos Area Codel Robert Steigerwald, Sr. Licensing Engineer (315) 349-6209 REPORT
CAUSE
SYSTEM COMPONENT MANUFACTURER
CAUSE
SYSTEM COMPONENT MANUFACTURER T EP gp SUPPLEMENTAL REPORT EXPECTED (14)
MONTH DAY YEAR EXPECTED YES SUBMISSION x
NO (If yes, complete EXPECTED SUBMISSION DATE).
DATE (15)
ABSTRACT (Limit to 1400 spaces, i.e., approximately 15 single-spaced typewntten hnes) (16)
On January 14, 1999, at 12:56 hours, with the plant operating at 100 percent power, a fire was reported in the Hydrogen (H Unusual Event Emergency Classification was ceclared, b)n Storage Facility. An s te and off-site fire fighting resources were dispatched.
The fire was declared extincJuished at 19:45 hours. The plant remained at 100 percent power for the duration of the event.
During the event, the plant's offsite 115 KV reserve power was de-energized to protect firefighters working in proximity to the 115 KV switchyard.
The Emergency Diesel Generator (EDG) Systems were started and run to demonstrate operability required by the plant's Technical Specifications.
During the Surveillance Test to assure operability, it was decided to leave the EDGs running unloaded.
However, the tie-breakers that force parallel the two EDGs on each Class IE emergency bus were open.
In this configuration the plant was in a condition that was outside the design basis of the plant and was required to be in cold shutdown within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
The causes for running the EDG Systems with the tie-breakers open were inadeguate procedures and insufficient operator training on the impact of EDG tie-breaker position on EDG system operability.
Corrective actions
include a revision to the EDG Operating and Surveillance Test Procedures and operator training program.
The most probable causes for the H fire were 2
a lack of oversight of the H vendors preventive maintenance program and 3
inadequate H system monitoring.
2 9904070181 990331 PDR ADOCK 05000333 0
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NaC FORM 366A U.S. NUCLEAR REGULATORY COMMISSION (6-1998)
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JImes A. FitzPatrick Nuclear Power Plant 05000333 2 OF 19 99 001 01 TEXT (11 more space et required, use addiconal copies of NRC Form 366A) (11}
EIIS Codes in []
Nste:
Immediately following the body of this LER are the following:
Figure 1,
" Simplified Power Distribution Sketch", Figure 2,
" Flow Diagram of H Storage Facility", and Attachment 1 which contains the timeline of 2
this event.
EVENT DESCRIPTION
On January 14, 1999 at 1256 hours0.0145 days <br />0.349 hours <br />0.00208 weeks <br />4.77908e-4 months <br />, with the plant operating at 100 percent rated power, a fire was reported in the plant's Hydrogen (H ) Storage 2
Facility [LJ) located within the northwest side of the protected area.
The report was made by an operator who had previously been at the H2 Storage Facility in preparation for putting the Hydrogen Injection System in service per Operating Procedure OP-89A.
The Hydrogen Storage Facility includes two high pressure storage banks and a pressure control station.
Each of the two high pressure storage banks (one is used for reserve) contains 15 storage vessels and have a maximum allowable working pressure of 2450 psig.
The pressure control station includes two full-flow pressure reducing regulators to maintain the required H Pressure for supporting plant systems.
Located in the 2
immediate vicinity of the H Storage Facility was a temporary hydrogen tank 2
truck trailer used for recharging the existing vessels.
This assembly was connected to the H storage facility; however, the supply line was isolated 2
at the trailer location.
The facility is located approximately 250 feet from the nearest safety-related structure (Diesel Generator Building) [NB].
Also near the facility is the 115 KV [FK) switchyard.
During this event, the reserve station 115KV [FK) lines were de-energized to protect the firefighters from electrocution while working around the 115KV switchyard.
Technical Specifications (TS) state that with both 115KV lines de-energized, continued reactor operation is permissible for a period not to exceed 7 days, provided that both redundant Emergency Diesel Generator Systems are operable, i
The JAF Emergency Power System [EK) is configured with two independent, fully redundant class IE power sources.
Each class IE power source is configured with two Emergency Diesel Generators (EDGs) which start, accelerate, and force parallel via a tie-breaker prior to " closing on the safety bus" in response to an automatic initiation signal.
Each class IE
]
power source receives an automatic initiation signal due to either a Loss of Coolant Accident (LOCA) or a degraded voltage condition on their respective safety bus.
lU.S. NUCLEAR REGULATORY COMMISSION l
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l J mes A. FitzPatrick Nuclear Power Plant 05000333 3 OF 19 99 001 01 TEXT (It more space is required, use additional copies of NRC form 366A) (17)
EVENT DESCRIPTION (cont'd.)
Note: The term " force parallel" means that the tie-breaker between an EDG pair (Breakers 10504 and 10604) shuts prior to electrically flashing the generators of the EDG pair.
Plant operators demonstrated operability of the EDGs by initiating Operations Surveillance Test Procedure ST-9D, "EDG, 115KV Reserve Power, Station Battery, or ESW Inoperable Test", which simulates an EDG start and force-parallel sequence.
While running this surveillance test, it was decided to continue to run (unloaded) the EDGs to cope with any potential loss of power transient that might develop if plant conditions were to degrade as a result of the H fire.
2 At this point in the event, plant operators were performing ST-9D, the EDGs were running unloaded and the EDGs were allowed to continue to run by stopping the execution of ST-9D and simply "not exiting" the procedure.
At this point in the procedure, all four EDGs (two per class IE power source) were running and their cross-tie breakers and output breakers were open.
On January 21, 1999 a post-event analysis of this configuration determined that, if the class IE power source (s) had received an automatic initiation signal at this time, protective relay logic would have only allowed one EDG per class IE power source to "close" on its respective safety bus.
This response is by design and is intended to protect an operating EDG from connecting to the safety bus out of phase synchronization.
j The protective relaying " checks" for this condition by monitoring the state of the cross-tie breaker.
Since the cross-tie breakers for both class IE power sources were open, phase synchronization could not be assured and each EDG would have entered a " relay race" to determine which EDG on each i
bus would "close" first.
The net result of this scenario would have been that one EDG rather than two (as required by design) would have energized each class IE power source.
JAF does not have an analvais which demonstrates that a single EDG on a safety bus can supply power to all loads assumed to be required to mitigate the design basis LOCA; therefore, this condition was determined to be outside the design basis of the plant.
The FitzPatrick Technical Specifications do not include a Limiting Condition for Operation Action Statement for this plant condition; therefore, this condition was also I
determined to be a condition prohibited by plant Technical Specifications.
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1 NhC FORM 366A t6-1938)
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,U.S. NUCLEAR REGULATORY COMMISSION (6 1998)
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James A. FitzPatrick Nuclear Power Plant 05000333 4 OF 19 99 001 01 TEXT (11 more space is required, use additional copies of NRC Form 366A) (17)
EVENT CAUSE - H FIRE 2
A root cause evaluation team, comprised of Power Authority, industry, and fire analysis personnel, has completed an in-depth investigation into the most probable cause of the fire event.
Following extensive equipment / component disassembly and evaluation, laboratory fracture analyses, and maintenance, operational, and design history reviews, the i
results indicate the following sequence of events as the most probable scenario for the H leak and ignition.
2 (Note:
Use Figure 2,
" Piping Diagram of H Storage Facility" as a valve 2
reference guide.)
The event was initiated as valve No. 212 (pressure control valve PCV-216 j
inlet isolation valve) was cracked open.
The operator performed two valve j
operations at the H Control Panel; first, valve No. 210 (pressure control j
2 valve PCV-215 inlet isolation valve) was opened, and second, valve No. 212 was cracked open (quarter of a turn).
Opening valve No. 210 had no effect on the system since the active regulator outlet isolation valve was shut and both H supplies (temporary H tank trailer and active high pressure H i
2 2
2 storage bank) were isolated.
The operator reported a popping noise and a panel fire immediately following operation of valve No. 212.
PCV-216 (reserve high-pressure storage bank pressure control valve) was found to have a ruptured diaphragm (most likely source of the popping noise).
The diaphragm failure was a time dependent failure as a result of fatigue and improper maintenance (missing a diaphragm disc).
This resulted in pressurizing the downstream low pressure piping to approximately 1800 psig.
i Valve No. 213 (PCV-216 outlet isolation valve) packing leak is the most probable initiator of the fire.
The configuration of the valve is such that a packing area leak is directed into the underside of a ductile iron valve handwheel.
Because of its low-ignition energy, when gaseous H is 2
released at high pressure, small heat producing sources (e.g.,
friction and static generation) often result in prompt ignition.
The high pressure H leak into a ductile iron handwheel was sufficient to ignite the leak.
2 Corrosion products on the valve internals indicate a history of moisture in the system which exacerbated the leak.
The ruptured diaphragm on PCV-216 provided a second H leakage path out the bonnet weep hole directed at 2
valve No. 213.
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J mes A. FitzPatrick Nuclear Power Plant 05000333 5 OF 19 99 001 01 i
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TEXT W rnore space is required, use additional copies of NRC Forrn 366A) (11)
EVENT CAUSE -H FIRE (cont'd.)
2
)
Valve RV-214 (low-pressure header relief valve) did not initially lift (frozen) based on the operator's account, but did lift later in the event j
(valve soft seat material found in discharge piping).
The sizing of relief 1
valve RV-214 is such that it cannot relieve the total high pressure flow from a failed open PCV-215 or 216.
Therefore, the H Control Pane.1 was at 2
relatively high pressures and temperatures for a length of time.
The identified failed equipment was a result of the original fire source.
Three separate failures occurred, or pre-existed at the initiation of the event (packing leak, PCV diaphragm, and RV freezing).
Each of the three initiators have previously occurred individually at FitzPatrick, and individually, these failures have never resulted in a fire.
When all occurred concurrently, a fire resulted.
The root cause determined that organizational and programmatic deficiencies.
j were present that led to the above failures.
The H panel and associated j
2 equipment are vendor supplied and vendor maintained.
The findings from this analysis identified two specific areas of concern.
Lack of Interface Reauirements:
An inadequate vendor maintained preventive maintenance program was being administered for the on-site H equipment.
2 Less than effective oversight of the vendor PM program was being maintained.
JAF standards for preventive maintenance had not been imposed on the vendor.
PM program requirements were not included in the initial contractual agreement with the vendor.
Inadeauate System Monitorina or Manaaement:
A review of the maintenance and operational history indicaLed a trend of continued system problems with less than effective engineering resolution of these performance issues.
EVENT CAUSE - EDG SYSTEMS OPERATED OUTSIDE DESIGN BASIS A heightened awareness of all plant parameters was being maintained during the fire event, however, Control Room operators did not recognize that running the EDG Systems in the unloaded condition with their respective tie-breakers open and not connected to the emergency bus as having the potential of rendering the two EDG systems inoperable in the event of a loss of power to the emergency busses.
The causes for this event were determined to be:
(1)
Inadeauate Procedures: Operating procedures provided no guidance or warning that EDG system operability was affected by EDG tie-breaker position.
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" ^a James A. FitzPatrick Nuclear Power Plant OtiOOO333 6 OF 194 001 01
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TEXT Uf more spsceis required, use additionalcopies of NRC Forrn 366A) (11)
EVENT CAUSE - EDG SYSTEMS OUTSIDE DESIGN BASTE (cont 'd, )
(2)
Operator training did not clearly describe the impact of EDG tie-breaker position on EDG system operability.
A contributing cause to this event was that station Administrative Procedure AP-02.06, " PROCEDURE USE AND ADHERENCE", allowed procedure deviations during emergencies with only Shift Manager approval.
Further review by the ERO or plant managcment may have prevented or identified this as.
.mproper operation.
ANALYSIS This report is being submitted in accordance with 10 CFR 50.73 (a) (2) (ii),
"Any event or condition that resulted in the nuclear power plant being in a condition that was outside the design basis of the plant", and 10 CFR 50.73 (a) (2) (i) (B), "Any operation or condition prohibited by the plant's Technical Specifications."
Additionally, the information contained in this report describing the conditions asecciated with the fire in the Hydrogen Facility is being provided in accordance with Licensae Event Report voluntary reporting criteria.
The H Facility was designed in accordance with Electrical Power Research 2
Institute (APR) report U-5283-S-A, Guidelines for Permanent Boiling Water Reactor (BAR) Hydrogen Water Chemistry Installation - 1987 Revision.
Construction to thsse Guidelines provides assuranen that neither normal operation nur failure of the on site hydrogen storage facility will adversely impact the function or reliability of cafety-related equipment.
The Hydrogen Storage Facility was designed and constructed to support the plant's Hydrogen Injection System and provide hydrogen for cooling of the Main Turbine Generator [T].
Immediately upon notification of the fire, Emergenc y Plan Implementing Procedure, EAR-3, " Fire" was implemented. Operators dispatched both on-site l and off-site firefighters in accordance with EAR-3 to cope with the event.
Firefighters controlled and eventually extinguished the fire by allowing the fire to deplete the hydrogen inventory, The 6afety significance of this event is addressed from three perspectives.
Industrial / Personnel Safety:
The individual involved in the initiating event received first degree burns i
i on his face.
These injuries were considered minor and the individual returned to work the same day.
An industrial safety analysis of this event indicated that the individual was, at one point, surrounded by flame but l
was protected by the fire ret ardant foul weather gear he was wearing at the time.
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R James A. FitzPatrick Nuclear Power Plant 05000333 7 OF 19 99 001 01 TEKT (11 more space is require $ use udevonalcopies of NRC Form 366A) (17)
ANALYSIS (cont'd.)
The firefighters involved in this event were working near the 115KV reserve station feeds.
As a precaution, it was requested that these lines be de-energized to protect the firefighters from electrocution.
Sofe Shutdown capability:
An Engineering assessment was completed of single Emergency Diesel Generator loading with a unit trip.
In the event of a Main Turbine trip (loss of on-site AC power) during the three and one half hours the EDGs were run with tie-breakers open, the single operable EDG on each bus was within its design capability to achieve and maintain a safe plant shutdown without offsite powar (no initiation of Emergency Core Cooling System loads required).
In addition, both 115KV lines were available with operator action and could have been restored if necessary, and the second EDG per system could have been manually loaded.
Therefore, this event was not safety significant from the perspective of plant safe shutdown capability.
J Postulated Accidents:
During this event, the plant was in an unanalyzed condition for approximately a four-hour period.
The probability of various LOCA events (as a function of break size) occurring during this time period is summarized below.
Probability of i - Break Size Effectiveness of High Pressure Occurrence During Injection Systems a Four-Hour Time Period Large HPCI [BJ) & RCIC [BN] not effective due to RPV 4.6 E -8 I
depressurization Intermediate May or may not be within the makeup
- 1. 4 E - 7 capacity of HPCI & RCIC Small Within makeup capacity of HPCI & RCIC 1.4 E -6 Small-small Essentially a leak inside containment, 1.4 E -5 within makeup capacity of HPCI & RCIC r
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James A. FitzPatrick Nuclear Power Plant 05000333 8 OF 19 99 001 01 TEXT fit more space os required, use additionalcopies of NRC Form 366A) (17)
ANALYSIS (cont'd.)
For the more probable, smaller break scenarios, core cooling could have been maintained using the HPCI & RCIC systems.
This would have provided ample time for operators to manually restore the 115KV reserve station feeds or manually place the out of service EDGs on their respective safety busses.
It is therefore reasonable to expect that, for the higher probability small break LOCA scenarios, operators would have been able to effectively mitigate the consequences of the event and no core damage would have occurred.
1 For the less probable, large and intermediate break scenarios, plant operators retained the ability to manually restore the 115KV reserve station feeds or manually place the out of service EDGs on their respective safety busses.
While it cannot be. demonstrated that the time required to take this manual action was within the time assumed for an EDG start and load sequence in the accident analysis, the time would have been short and i
power to the safety busses would have been restored within minutes.
The low probability of the large and intermediate break accidents, coupled with the ability to establish power to the emergency busses in a timely manner, reduces the safety significance of these events.
CORRECTIVE ACTIONS - EDG OPERATION 1.
Operations Surveillance Test Procedures ST-9BA, "EDG A AND C FULL LOAD TEST AND ESW PUMP OPERABILITY TEST" and ST-9BB, "EDG B AND D FULL LOAD TEST AND ESW PUMP OPERABILITY TEST", and ST-9D, "EDG, 115KV RESERVE POWER, STATION BATTERY, OR ESW SYSTEM INOPERABLE TEST" were revised to add precautionary statements that warn the operator of the Diesel Generators inoperable status while sonducting these STs, and provide further guidance on entering required LCO actions (24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> LCO).
1 2.
Operating Procedure OP-22, " DIESEL GENERATOR EMERGENCY POWER" was revised to add precautionary statements to alert operators that an EDG system is inoperable if operated in the single diesel mode, or when operating both EDGs unloaded with the EDG tie-breaker open.
3.
An entry was made into the Operations Department Night Orders to inform operating crews of the operational consequences of running the EDGs in the unloaded condition with tie-breakers open, the effects on system logic, and required Technical Specification actions during periods when EDGs are running in this configuration.
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i NRC FOJr. 366A U.S. NUCLEAR REGULATORY COMMISSION (6 1998) i UCENSEE EVENT REPORT (LER)
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James A. FitzPatrick Nuclear Power Plant 05000333 9 OF 19 99 001 01 l
TEXT (H more space is required, use additional copies of NRC Form 366A) (11)
CORRECTIVE ACTIONS - EDG OPERATION (cont'd.)
4.
The licensed operator initial training and continued training materials will be updated to include cautions on EDG inoperable status while conducting STs and guidance regarding i
Technical Specification requirements.
(Scheduled Completion Date: April 28, 1999) 5.
Administrative Procedure AP-02.06, " PROCEDURE USE AND ADHERENCE" will be revised to provide appropriate guidance for procedure deviations during emergency conditions.
The guidance will include the requirement for additional reviews by the Emergency Response Organization or the Plant Manager, as applicable.
(Scheduled Completion Date:
April 30, 1999) 6.
Operations Department shift personnel will review and be trained on the expectations associated with the revised AP-02.06.
(Scheduled Completion Date: June 30, 1999)
CORRECTIVE ACTIONS - H FIRE EVENT 2
1 1.
A Temporary Modification has been installed to provide interim make-up hydrogen to the Main Turbine Generator and provide make-up to the H Injection System.
2 2.
The equipment vendor for the H Addition System will be 2
provided guidelines of minimum expectations / requirements for inclusion into a preventive maintenance program that will be implemented at JAF on a periodic basis.
(Scheduled Completion Date: September 30, 1999) 3.
Guidance will be established for inspection of the H control 2
panel to identify and correct potential H ignition sources 2
(rust, debris). (Scheduled Completion Date: June 30, 1999) 4.
The temporary Hydrogen Addition System design has been evaluated to ensure that the lessons learned from the root cause analysis have been incorporated.
System improvements included in the temporary modification are: proper relief valve sizing to accommodate flow through a failed pressure control valve; improved pressure control valve design application; realigning the direction of the Ha panel to face away from the tube banks to prevent potential flame impingement; improve system vent pathways for more effective system purging following system maintenance or prolonged shutdown; a detailed grounding plan; and improvements in the H facility's access / egress points to ensure personnel safety 2
if normal paths are obstructed.
WRC FORM 366A (6-1998) l
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.U.S. NUCLEAR REGULATORY COMMIS$!ON (6-1998)
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YEAR SE AL J:mes A. FitzPatrick Nuclear Power Plant 05000333 10 OF 19 l
99 001 01 l
I TEXT lit more space os required, use additional copies of NRC Form 366 ) (17)
CORRECTIVE ACTIONS -H FIRE EVENT (cont'd.)
2 l
l 5.
Administrative Procedure AP-15.01, " PROCUREMENT OF MATERIALS i
AND SERVICES" and Design Control Manual Procedure DCM-3A,
" PREPARATION AND CONTROL OF TECHNICAL PROCUREMENT SPECIFICATIONS" will be revised to include the requirement to evaluate and specify that equipment procured by these procedures, and which remain vendor owned and maintained, require a preventive maintenance program that is consistent with FitzPatrick or industry recognized standards. This requirement shall define responsibilities and ownership for implementing the effectiveness of the program.
(Scheduled l
Completion Date:
May 31, 1999) 6.
An extent of condition review will be completed to identify all additional plant equipment that is vendor owned or maintained and whose failure could potentially endanger plant personnel and/or effect plant safety. (Scheduled Completion Date: May 31, 1999) 7.
System Engineering will develop and implement means for improving Hydrogen Addition System monitoring. (Scheduled Completion Date: May 31, 1999) 8.
Operating Procedure OP-89A, " HYDROGEN ADDITION SYSTEM" has been revised to incorporate detailed instructions for operating the system, including partial loading, long-term lay-up of components, and system purging requirements.
9.
Operations Department Standing Order ODSO-17, " AUXILIARY OPERATOR PLANT TOUR AND OPERATING LOG" will be revised to incorporate an additional log reading of the plant hydrogen supply pressure indicator 89A-PI-1113.
This reading is crucial during periods of plant shutdown or extended system lay-up to ensure the status of the low pressure piping system is monitored. Operator training will be conducted on the importance of readings regarding leak detection.
(Scheduled Completion Date: June 30, 1999) 10.
A new Industrial Safety Program Procedure ISPP-4.7. " Gaseous Hydrogen System", was developed and implemented to minimize the potential for fire, explosion, or personnel injury when working within or upon the Gaseous Hydrogen System.
Training was conducted on the new procedure prior to implementation.
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99 001 01 TEXT lit more since is required, use additional copies of NRC Form 366Al (17)
CORRECTIVE ACTIONS - Ha FIRE EVENT (cont'd.)
11.
A task analysis and appropriate training on the new H2 system for Operations personnel was completed.
In addition, the Operations Training Program Review Committee (TPRC) will review the results of root cause analysis conducted for the H 2 fire event and determine the necessary training on the operation and handling of H gas and the Hydrogen Addition 2
System. (Scheduled Completion Date: June 30, 1999)
ADDITIONAL INFORMATION
Previous Similar Events:
None Failed Components:
1.
Component Identification:
Reserve High Pressure Hydrogen Storage Bank Pressure Control Valve Component Mark Number:
089-PCV-216 Component Descriptian:
Manufacturer - CASHCO (C164)Model HP-8 3/4 inch Inlet pressure rating - up to 3000psig Outlet pressure - 325 to 750 psig 2.
Component Identificetion:
Hydrogen Supply Relief Valve Component Mark Numbe-89A-RV-214 Component Description:
Manufacturer - Anderson Greenwood (A415)
Model 81MB68-4 Size - 3/4 inch inlet X 1 inch outlet Pressure Range - 100psig to 1000psig 3.
Component Identification:
Hydrogen Storage Reserve Bank 89A-PCV-216 Outlet Isolation Valve i
Component Mark Number:
89A-H21mS-213 Component Description:
Manufacturer -REGO Model NUL9511L Size - 1 inch Maximum Working Pressure - 3500psig I
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.U.S. NUCLEAR REGULATORY COMMISSION 16-1998)
LICENSEE EVENT REPORT (LER)
TEXT CONTINUATION FACILirY NAME 11)
DOCKET (})
LER NUMBER (6)
PAGE (3) fR "A"
N R
NU James A. FitzPatrick Nuclear Power Plant 05000333 13 OF 19 99 001 00 TEXT (It more space is regswed, use addivonal copies of NRC Form 366A) 11 1)
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- U.S. NUCLEAR REGULATORY COMMISSION (6 1998)
LICENSEE EVENT REPORT (LER)
)
TEXT CONTINUATION FACILITY NAME (1)
DOCKET (2)
LER NUMBER (6)
~ PAGE (3)
"OusEE "duSEa" Jcmes A. FitzPatrick Nuclear Power Plant 05000333 14 OF 19 99 001 00 TEXT lif rnare space is required, use additional copies of NRC form 366A) (17)
~
Timeline Summary of Fire Event Severe weather conditions existed with heavy falling snow, seven to nine mile per hour winds, temperature of 3 degrees Fahrenheit (F) and a wind chill of minus ten degrees F.
January 14, 1999 12:56 An operator, while performing tasks at the Hydrogen Storage Facility, is involved in fire incident. The operator proceeds to the nearest communications station and contacts the Control Room to report the fire and request immediate dispatch of the Fire Brigade. Emergency Plan Implementing Procedure EAP-3, " Fire" was initiated.
12:59 Upon receipt of the fire report, the Control Room contacts the Oswego County E-911 Emergency Center for the dispatch of off-site fire fighting support. The Fire Brigade is simultaneously dispatched to the west side of the plant.
13:02 As a precaution, the Control Room orders an evacuation of the Turbine Building [NM], the Old Administrative Building, and the Administrative Building Annex. All are structures located in the proximity of the fire.
13:07 An immediate response and assessment by the Fire Brigade, Fire Protection and Safety Department personnel identify a large volume hydrogen fed fire of substantial magnitude in the vicinity of the H Facility's control station enclosure and in the rear of 2
the H trailer unit, with potential danger to H storage tanks 2
2 exposed to the high heat area.
This information is communicated to the Control Room. An immediate second alarm assignment is dispatched for additional off-site fire support. Initial strategic fire attack plans and risk issuea continue to be evaluated.
13:11 The Control Room declares an Unusual Event (UE) Emergency Classification at the plant with activation of the Technical Support Center (TSC) and Operations Support Center (OSC).
1 i
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e NRC F6RM 366A U.S. NUCLEAR REGULATORY COMMISSION 16 1998)
L UCENSEE EVENT REPORT (LER)
TEXT CONTINUATION FACILITY NAME (1)
DOCKET (2)
LER NUMBER (6)
PAGE (3)
YEAR SEO AL Rf S N James A. FitzPatrick Nuclear Power Plant 05000333 15 LF 19 99 N1 00 TEXT lit more space is required, use addition.of copies of NRc Form 366A) (11)
. Attachment 1 (cont'd.)
13:14 Offsite fire assistance begins arriving at the plant and is l
immediately escorted to the on-scene fire command. Current and l
projected tactical / strategic fire fighting objectives are l
reviewed including apparatus placement, water delivery sources, and flame extinguishment.
I 13:15 Both the plant's Electric and Diesel Fire Pumps [KP] are started L
for immediate availability upon demand.
13:23 The Fire Brigade contact the Control Roor. and requests the de-energizing of the overhead 115 KV transmission lines (off-site power source) as a safety precaution for personnel and fire fighting operations near or under the lines.
13:26-The 10022 and 10012 Breakers (associated with the 115 13:28 KV power sources) are opened.
13:37 Large volumes of water are applied to the fire and surrounding equipment from various staged water supply stations. Tactical plans are to continue to apply 1r ge caliber water streams to the exposed tank areas to provide and maintain impingement controls.
Fire Command advises the Control Room that extinguishing the fire is not a viable option at this time, and that water will continue to be apolied until such time that the product within the storage bank vessels bleeds down. Fire conditions are reported as stable and recommendations are made to prepare for long term continuous fire attack.
13:58 Fire Command reports that the fire is under control with three attack linea operational and working to protect exposed vessel and tank trailer areas.
]
15:00 The first Fire Command briefing is conducted to evaluate and assess fire fighting operations. A briefing is also conducted with the General Manager of Operations to provide an update and status of conditions A report is received that the H equipment 2
vendor is en route to the plant to provide technical support for fire termination and recovery activities. Large volumes of water are continuing to be directed to the H storage area. Weather 2
conditions continued to be severe. Firefighting personnel are being rotated into rehabilitation areas for observation of over-exposure to the weather conditions. Fire Command has instructed all stations to closely monitor the fire for any observed or recognized changes in propagation, magnitude, color, pressure release or any other occurrences that are noted changes in fire conditions..
.U.S. NUCLEAR REGULATORY COMMISSION 16-1998)
UCENSEE EVENT REPORT (LER)
TEXT CONTINUATION FACILITY NAME (1)
DOCKET (2)
LER NUMBER (6)
PAGE (3)
YEA 84 SE U L
E S N p
J:mes A. FitzPatrick Nuclear Power Plant 05000333 16 OF 19 99 001 00 TEXT Uf more space os required, use additional copies of NRC form 366A) (11) (cont'd.)
15:45 Fire suppression and control efforts continue. Incident conditions are being monitored for changes. Personnel continue rotation into rehabilitation areas due to weather exposure.
16:30 Fire Command briefing held. Strategic and tactical operations to continue as currently in place.
17:30 Fire command briefing held. Fire conditions are reported as improved with notable reduction in visible flames and associated pressure release of product. Utilizing binoculars, fire support personnel identify a flame source in the area of the H control 2
station and occasional evidence of small pockets of open flames from unidentified valve locations. A determination is made to reduce the amount of water flow to the fire area. Also, a determination is made to send a reconnaissance team closer to the fire proximity to try and assess the nature and source of the visible flame areas. Contingency plans are developed for manpower and command continuity for upwards of 12 additional hours.
18:00 The reconnaissance team is assembled and approaches the H2 facility. Observation team reporto that it appears that a single product line is the source of the open flames and is impinging in the direction or area of the H control station. The H equipment 2
2 vendor has r.rrived on site and is being escorted to the Fire Command station.
18:30 The fire brigade team is relieved by oncoming shift personnel.
Discussions are ongoing with the equipment vendor on tactical actions in support of extinguishing the remaining flames. Water flow from current operating mainstreams were directed to be shut down based upon diminishing fire conditions.
19:00 An approach to the immediate fire area is made by Fire Brigade personnel and H equipment vendor. An assessment and status of 2
the H Facility is performed and actions (manual valve 2
isolations) are completed to assure all vessel isolations including the trailer, are isolated. One small fire source remains, however, this eventually self-extinguishes. All high pressure storage cylinder bank pressures are at 0 pounds, except l
one reserve vessel, which has a reading of approximately 100 l
l pounds. Extensive damage to the storage facility is observed.
I
,U.S. NUCLEAR REGULATORY COMMISSION (6-1998)
UCENSEE EVENT REPORT (LER)
TEXT CONTINUATION FACILITY NAME (1)
DOCKET (2)
LER NilMBER (6)
PAGE (3)
NU NU R
James A. FitzPatrick Nuclear Power Plant 05000333 17 OF 19 99 001 00 TEXT Ilt more space is required, use additional copies of NRC Form 366A) (17) (cont'd.)
19:40 Off-site fire support demobilizes and exits the site.
The Fire Brigade has assumed full fire command.
19:45 The Fire Command reports to the Control Room that the fire is declared out and the H station is isolated.
2 21:03 The 115 KV off-site power lines are restored to service 21:05 Fire area is secured. Operations Department establishes a fire watch to monitor the area. The Emergency Director declares the Unusual Event terminated.
Summary of Operations Response to Fire Event January 14, 1999 12:56 A major fire is reported to the Control Room by a non-licensed operator who had previously been aligning valves at the Hydrogen Storage Facility in preparation for putting the Hydrogen Injection System in service. Fire Brigade is dispatched in accordance with Emergency Plan Implementing Procedure EAP-3,
" Fire".
13:02 The Control Room activates the station alarm. An evacuation of the Turbine Building, Old Administration Building and West Annex Building is ordered.
13:08 Control Room is notified that the fire plume is extending toward the temporary H tank trailer located adjacent to the H 2
2 facility. The Control Room is notified that off site fire assistance is proceeding to the plant. The Control Room is requested by the Fire Brigade leader to again notify Oswego County Fire Control and request a second alarm assignment be dispatched.
13:11 The Shift Manager (SM) declares a UE Emergency Classification at the James A.
FitzPatrick Nuclear Power Plant and assumes the position of Emergency Director (ED). The Technical Support Center (TSC) and Operations Support Center (OSC) are activated. However, alternate locations for the TSC and OSC are manned due to the Old Administration Building proximity to the normal TSC and OSC.
kRG FORM 366A 66-1998)
i OU.S. NUCLEAR REGULATORY COMMISSION LICENSEE EVENT REPORT (LER)
TEXT CONTINUATION
- FACluTY NAME (1)
DOCKET (2)
LER NUMBER (6)
PAGE (3)
Sus [f "EuS$
"^*
James A. FitzPatrick Nuclear Power Plant 05000333 18 OF 19 99 001 00 TEXT (11 more space as required, use additional copies of NRC Form 366A) (11) f (cont'd.)
13:15 At the request of the Fire Brigade Leader, the Control Room enters Operating Procedure OP-33, " Fire Protection" for initiation of both the Electric and Diesel Fire Pumps.
13:20 Via the Emergency Communications System (RECS) from the Control Room, New York State, Oswego County, and Nine Mile Point Nuclear Power stations (Units 1 and 2) are notified of the declaration of the UE at the plant.
13:20 The Site Executive Officer (SEO) relieves the SM of ED duties.
(Approx)
Due to the temporary relocation of the TSC, the ED remains located in the control Room for the duration of the UE.
13:23 The Control Room is requested by the Fire Brfgade to de-energize the overhead 115 KV reserve power lines as a safety precaution for fire fighting operations under and near the lines.
i 13:24 The Control Room notifies Nine Mile Point Unit 1 Nuclear Station and the state wide transmission station at Clark Energy Center that the 115 KV yard will be de-energized.
13:26-115 KV breakers are opened.
13:28 13:29 Fire Brigade is notified that the 115 KV overhead lines have been de-energized. Technical Specifications Limiting Conditions for Operation is entered for 115 KV lines out of service. (Note:
This LCO requires performance of Operations Surveillance Test Procedure ST-9D, "EDG, 115KV Reserve Power, Station Battery, or i
ESW System Inoperable Test")
13:30 The Control Room Supervisor (CRS) conducts crew briefings on j
conditions associated with de-energizing 115 KV reserve power lines. Preparations are made to start the four Emergency Diesel Generators and run them unloaded. Operator assignments are made for potential scram responsibilities. Discussions were conducted involving compensatory actions in the event of equipment failure.
13:36 The Control Room uses Operations Surveillance Test Procedure ST-9D to start EDGs A and 2.
Both EDGs are run without loading their respective generators.
NRC FORM 36bA (6-1998#
.U.S. NUCLEAR REGULATORY COMMISSION (6-1998)
LICENSEE EVENT REPORT (LER)
TEXT CONTINUATION FACILITY NAME (1)
DOCKET (2)
LER NUMBER (6)
PAGE (3)
YEAR SE
^L NU U
R J:mes A. FitzPatrick Nuclear Power Plant 05000333 19 OF 19 99 001 00 TEXT (11more space is required, use additional copies of NRC form 366A) (17} (cont'd.)
13:37 The Control Room is updated by the Fire Brigade on the status of the fire and fire fighting equipment, and that there exists no damage to the 115 KV lines due to their proximity to the fire.
13:38 The Control Room, via the Emergency Notification System (ENS)
)
telephone, notifies the NRC that a UE has been declared at the plant.
13:40 The Control Room starts and runs EDGs B and D without loading the respective generators.
13:41 The CRS conducts a crew briefing of plant status, crew responsibilities and plant response to potential power loss.
A standby operating crew arrives at the Control Room to provide assistance.
14:00-Control Room continues to monitor and receive updates 17:00 on fire status.
17:07-EDGs A, B,
C and D are secured.
17:08 i
19:45 The Fire is reported out at the H facility. Oswego County, Nine j
2 Mile Point Nuclear Stations (Units 1 and 2) and NRC are notified that the fire is out.
21:02 Control Room restores power to the 115 KV lines, the T.S. LCO for inoperability of the 115 KV is exited.
21:05 The Emergency Director secures from the Unusual Esent.
i