Information Notice 2000-01, Operational Issues Identified in Boiling Water Reactor Trip and Atransient
| ML003682692 | |
| Person / Time | |
|---|---|
| Issue date: | 02/11/2000 |
| From: | Marsh L Operational Experience and Non-Power Reactors Branch |
| To: | |
| Marsh L | |
| References | |
| IN-00-001 | |
| Download: ML003682692 (7) | |
UNITED STATES
NUCLEAR REGULATORY COMMISSION
OFFICE OF NUCLEAR REACTOR REGULATION
WASHINGTON, D. C. 20555-0001 February 11, 2000
NRC INFORMATION NOTICE 2000-01: OPERATIONAL ISSUES IDENTIFIED IN BOILING
WATER REACTOR TRIP AND TRANSIENT
Addressees
All holders of licenses for nuclear power reactors.
Purpose
The U.S. Nuclear Regulatory Commission (NRC) is issuing this information notice to alert
addressees to equipment and procedural issues experienced in a recent transient at the Hatch
nuclear power plant. It is expected that recipients will review the information for applicability to
their facilities and consider actions, as appropriate, to avoid problems. However, suggestions
contained in this information notice are not NRC requirements; therefore, no specific actions or
written response is required.
Description of Circumstances
On January 26, 2000, at Hatch Unit 1, the reactor automatically scrammed on low reactor water
level after a partial loss of feedwater occurred. One of two main feedwater lines was isolated
when a valve unexpectedly closed in the feedwater flow path to the reactor. The licensee later
determined that the valve closed because of a problem with the valve control switch. As a
result of the valve closure, feedwater flow was significantly decreased; therefore, reactor water
level decreased, and the reactor automatically scrammed as expected.
The high-pressure coolant injection (HPCI) system and the reactor core isolation cooling (RCIC)
system automatically actuated and injected water into the reactor as designed. These systems, along with the feedwater system, increased reactor water level rapidly. The feedwater and
RCIC systems tripped on high level as expected. However, the HPCI system did not
immediately trip as designed on high level and continued to inject water into the reactor for
about 1 minute before tripping. Reactor water level increased to the point that water entered
the main steam lines. The licensee closed the main steam isolation valves (MSIVs) in
accordance with the emergency operating procedure.
Pressure in the shutdown reactor began to slowly increase because of decay heat. A licensee
operator attempted to open a safety relief valve to control reactor pressure but did not receive
the expected indications on the control panel. The operator then actuated the control switches
for other safety relief valves until he received the expected open indication on one valve.
Subsequently, several safety relief valves were operated satisfactorily to control reactor
pressure. Later, the licensee determined that the safety relief valves had opened properly
when actuated. Safety relief valve tailpipe temperature indications, available on a control room
-2- back panel recorder, clearly indicated the valves had operated. Reactor pressure reached a
maximum value slightly above normal operating pressure and did not approach an operational
safety limit.
The licensee controlled the reactor water level using HPCI and RCIC. Although initial attempts
to restart RCIC were unsuccessful, the licensee was able to use the system later in the event.
HPCI was manually operated several times for water level control and the licensee observed
that it tripped properly at the high-level setpoint twice during the recovery.
On January 30-February 5, 2000, the NRC conducted an augmented team inspection (AIT) of
the circumstances of this event. The objectives of this inspection were to (1) determine the
facts of the event, (2) assess the licensees response to the event, (3) assess the licensees
event review and recovery actions, and (4) assess any generic aspects of the event.
Discussion
In this event, several systems did not perform as expected.
Safety Relief Valves
The licensees investigation into the response of the safety relief valves focused on the valves
position indication, the effect water has on the operation of the safety relief valve, and the effect
that water passing through the safety relief valve has on the tailpipes, tailpipe vacuum breakers, and tailpipe pressure switches. The licensee was assisted by the nuclear steam supply system
vendor, General Electric (GE), and the safety relief valve vendor, Target-Rock, in conducting
this investigation and assessment. The licensee concluded that the safety relief valves
operated each time the control switches were actuated in the control room. However, the
operators were unaware that the safety relief valves were open because they did not receive
the expected indicating light on the control panel. A pressure switch located in each safety
relief valve tailpipe actuates due to increased tailpipe pressure when the safety relief valve is
opened and, in turn, actuates an indicating light on the control panel. During this event, pressure in the tailpipes did not increase sufficiently to actuate the pressure switches while the
safety relief valve was passing water. The licensee sent several of the safety relief valve
control assemblies (topworks) to a valve test facility for testing and inspection. No
abnormalities as a result of this event were identified. The licensee conducted inspections of
the safety relief valve tailpipes and other plant components that may have been subjected to
the water in the steam lines and did not identify any adverse conditions that resulted from this
event.
Reactor Core Isolation Cooling
During the event, the RCIC system automatically initiated on low reactor water level and
continued to inject until the RCIC turbine steam admission valve closed on high reactor water
level as designed. Initial attempts to restart RCIC were unsuccessful. Water from the main
steam lines had entered the line supplying steam to the RCIC turbine, which affected the
turbine control system and resulted in closing the trip and throttle valve. The licensee
concluded that the closure of the trip and throttle valve was due to an electrical overspeed
-3- condition caused by water carryover into the turbine governor valve. In addition, the licensees
procedural guidance and training for restarting the tripped system with water in the steam
supply line was inadequate. The licensee successfully manually started the system later in the
event and identified no further problems with its operation.
The licensee concluded that, in accordance with its procedures and training, operators
attempted to restart the turbine by resetting and opening the trip and throttle valve with (1) the
steam admission valve full open and (2) the turbine control system demanding maximum
speed. This method results in rapid admission of steam into the turbine, which increases the
possibility of tripping the turbine on electrical overspeed. Additionally, it was determined that
the trip and throttle valve response on the simulator did not accurately model the actual valve
response in the plant.
The NRC has issued several information notices (listed below) on experiences at other nuclear
power plants with water in the steam supply to turbine-driven pumps.
High Pressure Coolant Injection
Early in the transient, the system initiated, as designed, upon reaching its reactor water
low-level setpoint and injected to assist the RCIC and feedwater systems in recovering the
reactor water level. The system did not immediately trip upon reaching the high reactor water
level but tripped after about 1 minute of continued operation. Later in the transient, the licensee
manually restarted HPCI several times for reactor water level and pressure control. The system
promptly tripped, as designed, at the high-level setpoint on two occasions.
The licensee conducted a detailed investigation regarding HPCI operation and did not
conclusively determine why the system did not immediately trip during its initial operation.
Testing of the associated components failed to identify the cause of the event but supported
operability of the system.
Feedwater Valve Handswitches
The partial loss of feedwater occurred when a valve in the main feedwater flow path to the
reactor closed unexpectedly. Later, the licensee determined that the valve closed because of
a malfunction of a GE Type CR 2940 control switch. In 1977, GE issued Service Information
Letter No. 217, which indicated that this model control switch was overly sensitive during
positioning and that the switch contacts may close prematurely from the slightest movement of
the selector switch.
Performance of Licensed Operators:
Several operational performance issues complicated the transient and recovery. For example, after the initial injection, several efforts to restart RCIC were unsuccessful because the
procedural guidance and simulator training were not adequate for the existing conditions. The
event occurred during shift turnover when a large number of operators were in the control room
and resulted in unclear lines of responsibility and communication difficulties during some
phases of the event. For example, there was a slight delay by the operators in shutting the
-4- MSIVs. Additionally, the operators did not identify that HPCI did not immediately trip at the
high-level setpoint.
Health and Safety Assessment
The AIT concluded that the event did not adversely affect the health and safety of the public.
The event did not result in a radiological release, and no operational safety limits were
approached. Safety-related systems remained capable of accomplishing their required safety
functions, although some problems occurred with important plant equipment. No need existed
to declare an unusual or emergency condition.
Generic Implications
The AIT concluded that several issues identified during the inspection potentially have generic
implications. They are:
1. Safety relief valve operation and indication is affected when the valve is passing water
instead of steam. Opening times may be slower, on the order of several seconds
versus milliseconds. Tailpipe pressure experienced when passing water may not be
sufficient to actuate pressure switches used for position indication.
2. Procedural guidance for closing the main steam isolation valves and setpoints for the
high-level trips of the injections systems may not prevent complications due to water
collecting in the main steam lines.
3. RCIC performance is affected by resetting the turbine trip and throttle valve with the
steam admission valve open and a flow demand present, especially if excessive
moisture is present in the steam supply to the turbine.
Related Generic Communications
C Information Notice 85-50, Complete Loss of Main and Auxiliary Feedwater at a PWR
Designed by Babcock & Wilcox, July 8, 1985 C Information Notice 85-76, Recent Water Hammer Events, September 19, 1985 C Information Notice 86-14, PWR Auxiliary Feedwater Pump Turbine Control Problems, March 10, 1986 C Information Notice 86-14, Supplement 1, Overspeed Trips of AFW, HPCI, and RCIC
Turbines, December 17, 1986 C Information Notice 86-14, Supplement 2, Overspeed Trips of AFW, HPCI, and RCIC
Turbines, August 26, 1991 C Information Notice 88-77, Inadvertent Reactor Vessel Overfill, December 17, 1986
-5- This information notice requires no specific action or written response. If you have any
questions about the information in this notice, please contact one of the technical contacts
listed below or the appropriate Office of Nuclear Reactor Regulation (NRR) project manager.
/RA J. E. Lyons
Acting for Ledyard B. Marsh, Chief
Events Assessment, Generic Communications
and Non-Power Reactors Branch
Division of Regulatory Improvement Programs
Office of Nuclear Reactor Regulation
Technical contacts: Vern Hodge, NRR Len Wert, Region II
301-415-1861 404-562-4540
E-mail: cvh@nrc.gov E-mail: lxw1@nrc.gov
Dave Skeen, NRR Joel Munday, Region II
301-415-1174 404-562-4520
E-mail: dls@nrc.gov E-mail: jtm@nrc.gov
Attachment: List of Recently Issued NRC Information Notices
-5- This information notice requires no specific action or written response. If you have any
questions about the information in this notice, please contact one of the technical contacts
listed below or the appropriate Office of Nuclear Reactor Regulation (NRR) project manager.
/RA J.E. Lyons
Acting for Ledyard B. Marsh, Chief
Events Assessment, Generic Communications
and Non-Power Reactors Branch
Division of Regulatory Improvement Programs
Office of Nuclear Reactor Regulation
Technical contacts: Vern Hodge, NRR Len Wert, Region II
301-415-1861 404-562-4540
E-mail: cvh@nrc.gov E-mail: lxw1@nrc.gov
Dave Skeen, NRR Joel Munday, Region II
301-415-1174 404-562-4520
E-mail: dls@nrc.gov E-mail: jtm@nrc.gov
Attachment: List of Recently Issued NRC Information Notices
DOCUMENT NAME: C:\in20001.wpd
- See previous concurrence
To receive a copy of this document, indicate in the box C=Copy w/o attachment/enclosure E=Copy with attachment/enclosure N = No copy
OFFICE PECB Tech Editor D:DRP C:PECB
NAME CVHodge* BACalure* LRPlisco*email LBMarsh/JELyons for
DATE 02/10/00 02/10/00 02/11/00 02/11/00
OFFICIAL RECORD COPY
Attachment LIST OF RECENTLY ISSUED
NRC INFORMATION NOTICES
_____________________________________________________________________________________
Information Date of
Notice No. Subject Issuance Issued to
_____________________________________________________________________________________
99-34 Potential Fire Hazard in the 12/28/99 All holders of licenses for nuclear
use of Polyalphaolefin in reactors and fuel cycle facilities
Testing of Air Filters
99-33 Management of Wastes 12/28/99 All medical licensees
Contaminated With
Radioactive Materials
99-32 The Effect of the Year 2000 12/17 All NRC medical licensees
Issues on Medical Licensees
99-31 Operational Controls to Guard 11/17/99 All NRC licensed fuel cycle
Against Inadvertent Nuclear conversion, enrichment and
Criticality fabrication facilities
99-30 Failure of Double Contingency 11/8/99 All fuel cycle licensees and
Based on Administrative certificates performing laboratory
Controls Involving Laboratory analysis to determine uranium
Sampling and Spectroscopic content, in support of
Analysis of Wet Uranium administrative criticality safety
Waste controls
99-29 Authorized Contents of Spent 10/28/99 All power reactor licensees and
Fuel Casks spent fuel storage licensees and
applicants
99-01, Rev. 1 Degradation of Prestressing 10/7/99 All holders of operating licensees
Tendon Systems in for nuclear power reactors
Prestressed Concrete
Constrainments
99-28 Recall of Star Brand Fire 9/30/99 All holders of licenses for nuclear
Protection Sprinkler Heads power, research and test
reactors, and fuel cycle facilities
99-27 Malfunction of Source 9/2/99 All medical licensees authorized
Retraction Mechanism in to conduct teletherapy treatments
Cobalt-60 Teletherapy
Treatment Units
____________________________________________________________________________________
OL = Operating License
CP = Construction Permit
____________________________________________________________________________________
OL = Operating License
CP = Construction Permit