Information Notice 2009-22, Recent Human Performance Issues at Nuclear Power Plants
| ML091940257 | |
| Person / Time | |
|---|---|
| Issue date: | 10/02/2009 |
| From: | Mcginty T Division of Policy and Rulemaking |
| To: | |
| Eric Thomas, NRR/DIRS, 301-415-6772 | |
| References | |
| IN-09-022 | |
| Download: ML091940257 (6) | |
ML091940257 UNITED STATES
NUCLEAR REGULATORY COMMISSION
OFFICE OF NUCLEAR REACTOR REGULATION
WASHINGTON, DC 20555-0001
October 2, 2009
NRC INFORMATION NOTICE 2009-22:
RECENT HUMAN PERFORMANCE ISSUES AT
NUCLEAR POWER PLANTS
ADDRESSEES
All holders of operating licenses or construction permits for nuclear power reactors under the
provisions of Title 10 of the Code of Federal Regulations (10 CFR) Part 50, Domestic Licensing
of Production and Utilization Facilities, except those who have permanently ceased operations
and have certified that fuel has been permanently removed from the reactor.
PURPOSE
The U.S. Nuclear Regulatory Commission (NRC) is issuing this information notice (IN) to inform
addressees of recent plant events that were attributed at least in part to human performance
issues. The NRC expects recipients to review the information for applicability to their facilities
and to consider actions, as appropriate, to avoid similar problems. Suggestions contained in
this IN are not NRC requirements; therefore, no specific action or written response is required.
BACKGROUND
The NRC is issuing this IN to communicate the circumstances, details, and consequences of
several recent events in the cross-cutting area of Human Performance. Previously issued INs
related to similar plant events include the following:
IN 2005-16, Outage Planning and Scheduling - Impacts on Risk, dated June 20, 2005
IN 2007-11, Recent Operator Performance Issues at Nuclear Power Plants, dated
March 6, 2007
In addition, NUREG-1449, Shutdown and Low Power Operation at Commercial Nuclear Power
Plants in the United States, dated September 1993 highlights several of the same event types
discussed in this IN. These include loss of reactor coolant inventory, pressurized-water reactor
venting, and inadvertent reactivity addition.
DESCRIPTION OF CIRCUMSTANCES
Oconee Nuclear Station, Unit 1
On April 12, 2008, while performing plant cooldown and depressurization activities for a
refueling outage at Oconee Nuclear Station, Unit 1, operations personnel identified abnormally high vibrations on all three operating reactor coolant pumps (RCPs); the fourth RCP had been
secured before the plant was shut down. During the event, the licensee operated RCP 1A2 outside of its normal vibration tolerances to facilitate the cooldown and depressurization of the
reactor coolant system (RCS). The original purpose of operating the RCP at vibration levels in
excess of normal tolerances was to allow the initiation of low-pressure injection decay heat
removal. Licensee management approved this action after consulting with the pump vendor.
However, the pump remained in operation for an additional 2.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> after initiation of decay
heat removal, which eventually led to multiple seal failures on RCP 1A2 and an 8-gallon-per- minute RCS leak past the pump seals.
As a result of the RCP 1A2 seal failures and the subsequent RCS leak, the reactor building
iodine detector began to alarm. However, the operating crew did not recognize the condition, the alarm response procedure was not entered until 9 hours1.041667e-4 days <br />0.0025 hours <br />1.488095e-5 weeks <br />3.4245e-6 months <br /> later, and outage work, including
the removal of the reactor building equipment hatch, was allowed to continue. After securing
RCP 1A2, licensee management met and determined that RCP 1B2 should be restarted with
increased vibration tolerances to assist with RCS degasification and crud burst mitigation.
Within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> of starting RCP 1B2, two of its three seals failed, and the pump was secured. In
its response to the failed seals on RCP 1B2, the operating crew did not properly monitor plant
parameters as required by procedure. Adherence to the procedure would have alerted the crew
to the alarming reactor building iodine detector and to the severity of the ongoing RCS leak
caused by the RCP 1A2 seal failure. Additional details of this event can be found in NRC
Special Inspection Report 05000269/2008008, Agencywide Documents and Access
Management System (ADAMS) Accession No. ML082140807.
Oconee Nuclear Station took the following corrective actions related to this event:
identified vulnerabilities and improvements needed in RCP motor and seal design, monitoring, maintenance practices, spare parts, procedures, knowledge and expertise
presented an RCP case study to industry leadership academies
conducted training for plant management on the importance of maintaining
independence and focusing on equipment qualification, validation and verification when
performing its operational decisionmaking role
reviewed applicable abnormal operating procedures for entry conditions and
appropriateness for shutdown events
ensured clear criteria were established for containment closure and replacement of the
equipment hatch
changed plant procedures to emphasize the primary supervisory and oversight function
of the control room supervisor and to remove administrative burden Three Mile Island Nuclear Generating Station, Unit 1
During the Three Mile Island Unit Nuclear Generating Station refueling outage that began
October 22, 2007, there were several human performance-related issues that resulted in NRC
inspection findings which are summarized below, and in NRC Integrated Inspection Report 05000289/2007005, ADAMS Accession No. ML080430321.
TMIs procedure for authorizing workers to exceed work hour limits of 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />s/week
was determined to be deficient in that it did not provide reasonable assurance that
station management would properly control overtime for plant staff performing safety- related functions.
Due to a lack of coordination of maintenance and operational activities, installation of a
once-through-steam-generator (OTSG) primary lower manway cover during mid-loop
operation resulted in an unexpected drop in reactor vessel level indication. The change
in level indication prompted operators to declare an Unusual Event for the plant.
Reactor vessel level indication dropped because the OTSG lower manway cover was
installed while temporary ventilation fans were exhausting air from the OTSG handhole
RCS vent. This event resulted in an inspection finding with a cross-cutting aspect in
Human Performance Work Control because activities were not properly coordinated
while the plant was in an elevated shutdown risk condition.
Fuel handling operators failed to properly implement procedures while moving a control
rod assembly (CRA) in the spent fuel pool. After the grappling tool indicated that it had
failed to engage a CRA, operators moved the refueling bridge to the next fuel assembly
without first visually checking to ensure the CRA was not grappled. Oncoming fuel
handling operators visually checked the mast and noticed that the CRA was grappled.
Further investigation revealed that the CRA fingers were still inserted in the previous fuel
assembly and that the subsequent movement of the refueling bridge severely bent them.
Vogtle Electric Generating Plant, Unit 1
On August 19, 2008, the licensee was conducting a flush of the boron thermal regeneration
system (BTRS) chilled water piping. The procedure for this evolution requires that components
be manipulated both locally and in the control room. A system operator was directing the flush
from outside the control room, where most of the procedure steps for the flush take place. The
system operator read a procedure step to the Operator at the Controls (OATC) that required the
OATC to place a control room handswitch in the Closed position. Despite a requirement to do
so, the OATC did not have the procedure in hand when performing this evolution. Due to
misinterpretation of the procedure step, the OATC verified that the valve was closed but did not
place the handswitch in the Closed position.
When operators performed subsequent steps in the procedure, the valve automatically opened, and approximately 500 gallons of un-borated water were flushed from the BTRS to the volume
control tank. The resulting positive reactivity transient caused control rods to automatically
insert in order to compensate. Control room operators also responded to the transient by
continuing to insert control rods, borating, and reducing main generator load; but not before the
1-minute average reactor power level increased to 100.73 percent. Additional details are available in NRC Integrated Inspection Report 05000424/2008004, ADAMS Accession No.
Salem Nuclear Power Plant, Unit 1
On October 15, 2008, one day after shutting down for a refueling outage, control room operators
at Salem Nuclear Power Plant, Unit 1, began draining the pressurizer from a water-solid
condition to a target level of 10 to 15 percent. A higher-than-normal dissolved gas
concentration in the RCS, combined with a rapid decrease in RCS pressure, caused the
reference leg for the pressurizer cold-calibrated level instrument to partially void, resulting in an
erroneously high pressurizer level reading. Control room operators were not aware that the
reference leg was voiding and began the pressurizer drain down evolution using the cold- calibrated level instrument as their primary level indication.
Nearly 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> into the evolution, operators noted that the pressurizer level unexpectedly
stabilized at 80 percent. The control room supervisor directed the reactor operator to raise the
charging flow and increase RCS inventory. After troubleshooting the level instrument, technicians backfilled the reference leg of the pressurizer level cold-calibrated instrument, and
stabilized the level at 14 percent several hours after increasing the charging flow. The licensee
determined that the pressurizer had completely drained during this event allowing nitrogen from
an ongoing pressurizer relief tank purge to migrate into the top of the reactor vessel and into
some of the steam generator u-tubes. Additional details of this event can be found in NRC
Special Inspection Report 05000272/2008009, ADAMS Accession No. ML090200076.
The licensees investigation revealed several missed opportunities for mitigating this event, including the following:
The procedure for draining the pressurizer did not require operators to use diverse and
redundant indications for controlling RCS inventory.
When the Pressurizer Heater Off Level Low annunciator alarmed, operators did not
properly implement the response procedure, which could have alerted them to the
erroneous level indication sooner.
The licensee did not properly implement corrective actions for a similar industry event
that occurred in 1997 at the Sequoyah Nuclear Plant.
The licensee developed several corrective actions following this event, including procedure
updates to include diverse and redundant indications for RCS inventory monitoring, and
simulator training prior to outages that refreshes operator knowledge on upcoming evolutions.
DISCUSSION
The events related to human performance discussed above occurred, for the most part, during
refueling outages. They often involved either the failure of licensees to provide adequate
procedures or the failure of operators to properly follow those procedures. In several instances, it appeared that shutdown risk was either under-estimated or not considered. Overall, the
events involved behaviors that are often associated with an operators loss of safety focus, such as taking actions when the consequences of those actions are uncertain, or taking actions
outside the scope of the relevant procedure(s). Monitoring and controlling reactivity and RCS
inventory in accordance with plant procedures are two of the most important responsibilities of
an on-duty licensed reactor operator. Operating experience shows that errors often occur
before, during, and immediately after refueling outages, when evolutions are less familiar to the
operators because they are infrequently performed. From a risk perspective, refueling outages
can constitute as much as one-third of a plants overall core damage frequency even though the
average plant is only in refueling outages about 5 percent of the time.
CONTACT
This IN requires no specific action or written response. Please direct any questions about this
matter to the technical contact listed below or to the appropriate Office of Nuclear Reactor
Regulation project manager.
/RA/
Timothy J. McGinty, Director
Division of Policy and Rulemaking
Office of Nuclear Reactor Regulation
Technical Contact:
301-415-6772 Eric.Thomas@nrc.gov
Note: NRC generic communications may be found on the NRC public Web site, http://www.nrc.gov, under Electronic Reading Room/Document Collections.
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