Information Notice 2011-02, Operator Performance Issues Involving Reactivity Management at Nuclear Power Plants: Difference between revisions
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| issue date = 01/31/2011 | | issue date = 01/31/2011 | ||
| title = Operator Performance Issues Involving Reactivity Management at Nuclear Power Plants | | title = Operator Performance Issues Involving Reactivity Management at Nuclear Power Plants | ||
| author name = Blount T | | author name = Blount T | ||
| author affiliation = NRC/NRR/DPR | | author affiliation = NRC/NRR/DPR | ||
| addressee name = | | addressee name = | ||
Line 13: | Line 13: | ||
| document type = NRC Information Notice | | document type = NRC Information Notice | ||
| page count = 6 | | page count = 6 | ||
}} | }} | ||
{{#Wiki_filter: | {{#Wiki_filter:UNITED STATES | ||
NUCLEAR REGULATORY COMMISSION | |||
OFFICE OF NUCLEAR REACTOR REGULATION | |||
WASHINGTON, DC 20555-001 January 31, 2011 NRC INFORMATION NOTICE 2011-02: OPERATOR PERFORMANCE ISSUES | |||
INVOLVING REACTIVITY MANAGEMENT AT | |||
NUCLEAR POWER PLANTS | |||
==ADDRESSEES== | ==ADDRESSEES== | ||
All holders of operating licenses for nuclear power reactors under the provisions of Title 10 of the Code of Federal Regulations (10 CFR) Part 50, | All holders of operating licenses 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== | ==PURPOSE== | ||
The U.S. Nuclear Regulatory Commission (NRC) is issuing this information notice (IN) to inform addressees of events involving deficiencies with reactivity management planning and | The U.S. Nuclear Regulatory Commission (NRC) is issuing this information notice (IN) to inform | ||
addressees of events involving deficiencies with reactivity management planning and | |||
implementation. 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. | |||
==DESCRIPTION OF CIRCUMSTANCES== | ==DESCRIPTION OF CIRCUMSTANCES== | ||
Callaway Plant During a Callaway Plant shutdown in October 2003, the control room operators did not effectively control core reactivity during low-power | |||
===Callaway Plant=== | |||
During a Callaway Plant shutdown in October 2003, the control room operators did not | |||
effectively control core reactivity during low-power operations. The event began on the morning | |||
of October 20, 2003, when the Callaway Plant experienced an inverter failure on a | |||
safety-related bus that put the unit in a 24-hour technical specification action to restore the | |||
inverter or be in Mode 3 (hot standby) within the next 6 hours. The next morning, because the | |||
inverter had not yet been restored, operators initiated a plant shutdown at approximately | |||
10 percent per hour. With the main turbine on line and with turbine bypass valves closed, operators attempted to stabilize the plant at approximately 8-percent power. Per procedure, operators opened the turbine drain valves, which increased main steam flow, reducing reactor | |||
coolant temperature and adding positive reactivity. At the same time, negative reactivity was | |||
being inserted by xenon buildup, decreasing reactor power. The net effect was that reactor | |||
coolant temperature decreased by approximately 10 degrees Fahrenheit over a half-hour | |||
period. Operators did not withdraw control rods or dilute boron concentration to stabilize reactor | |||
power or reactor coolant temperature. As a result of the lowering reactor coolant temperature, pressurizer level lowered enough to cause letdown to isolate. In addition, reactor coolant | |||
temperature decreased below the technical specification required minimum temperature while | |||
critical for several minutes. In response to the lowering reactor coolant temperature, the | |||
operators manually tripped the main turbine with power at approximately 5 percent. Upon | |||
tripping the main turbine, the turbine bypass valves controlled steam pressure, causing reactor | |||
coolant temperature and pressurizer level to return to normal. Tripping the main turbine | |||
reduced main steam flow, increasing reactor coolant temperature and adding negative | |||
reactivity, which together with the addition of negative reactivity by xenon buildup, caused the | |||
reactor to become subcritical. Operators did not insert the control rods until almost 2 hours after | |||
the reactor became subcritical. | |||
A subsequent review of this plant shutdown found that control room operators did not effectively | |||
control reactivity to maintain the reactor in the desired condition during low-power operations by | |||
properly anticipating, controlling, and responding to changing plant parameters. Operators did | |||
not use control rods or boron concentrationtwo means that operators can use to directly | |||
control the amount and timing of reactivity changesto adjust for reactivity changes by xenon | |||
buildup and reactor coolant temperature changes. Specifically, reactivity was not effectively | |||
controlled in that (1) operators did not sufficiently anticipate and compensate for xenon buildup | |||
when they attempted to stabilize and hold the plant at approximately 8 percent power which | |||
caused reactor coolant temperature to continue to decrease below the technical specification | |||
required minimum temperature for criticality; (2) operators did not shut down the reactor in a | |||
deliberate manner (e.g., by inserting control rod banks), but rather the reactor became | |||
subcritical by xenon buildup and by the increase in reactor coolant temperature resulting from | |||
the operators manually tripping the main turbine; and, (3) operators did not insert control rods | |||
for nearly 2 hours after the reactor became subcritical. Fully inserting control rods provides | |||
assurance that the reactor remains shut down (regardless of reactor coolant temperature or | |||
xenon concentration). Notwithstanding, the NRC determined the delay in completing the | |||
shutdown by inserting the control rods did not result in an unsafe reactor condition. | |||
Operator performance in not effectively controlling reactivity was attributable, in part, to | |||
weaknesses with management oversight, training, and procedural guidance. The pre-evolution | |||
practice session using the simulator did not cover plant operations below 10-percent power and | |||
did not include operation after the point where the operators tripped the main turbine. The | |||
reactivity management plan did not address the possibility that the expected reactivity change | |||
from tripping the main turbine together with the xenon buildup could cause the reactor to | |||
become subcritical. The licensees initial post-shutdown review did not identify and evaluate the | |||
atypical manner that the reactor became subcritical. This omission delayed application of the | |||
lessons learned to operator qualification and requalification training and significantly delayed | |||
procedure changes to address weaknesses in operator control of reactivity during low-power | |||
operation. This event is also discussed in Callaway PlantNRC Integrated Inspection Report | |||
05000483/2007003, dated August 2, 2007, which can be found on the NRCs public Web site in | |||
the Agencywide Documents Access and Management System (ADAMS) under Accession | |||
No. ML072140876. | |||
===River Bend Station=== | |||
On March 8, 2008, with River Bend Station at 25-percent power, control room operators were | |||
withdrawing control rods to increase reactor power. The operating procedure for plant startup | |||
directs operators to withdraw control rods using a withdrawal sequence specified in a reactivity | |||
control plan that is provided to them by licensee reactor engineering. However, the dedicated | |||
reactor operator at the controls stated an incorrect target position when reading aloud a rod movement step in the reactivity control plan. As a result, this operator individually withdrew six | |||
consecutive rods to position 24 rather than the target position 20 specified in the reactivity | |||
control plan. The dedicated peer-check reactor operator did not identify that the stated target | |||
position was incorrect because he could not readily see the reactivity control plan that was | |||
resting on the lap of the reactor operator at the controls. The operator at the controls halted the | |||
withdrawal of the seventh rod at position 18 after the dedicated peer-check reactor operator | |||
identified the error. The licensee determined that the reactor operator at the controls and the | |||
peer-checker did not follow the procedures to prevent human performance errors and that the | |||
senior reactor operator did not maintain effective oversight of the activity. Additional information | |||
is available in River Bend StationNRC Integrated Inspection Report 05000458/2008002, dated May 9, 2008 (ADAMS Accession No. ML081300838). | |||
===Diablo Canyon Power Plant, Unit 2=== | |||
In August 2009, Diablo Canyon Power Plant Unit 2 was shut down in order to troubleshoot and | |||
repair a main transformer bushing. In preparation for the shutdown, the control room operators | |||
performed simulator training on a ramp downpower using a draft copy of a ramp plan provided | |||
via e-mail by reactor engineering. Before the actual shutdown, a revised ramp plan was | |||
provided by reactor engineering, approved by the operations manager, and issued in the shift | |||
orders. This revised ramp plan was also e-mailed to all shift members. The oncoming shift | |||
foreman and shift manager did not review the approved ramp plan located in the shift orders nor | |||
did they review the ramp plan as part of the reactivity brief. Operators began the ramp | |||
downpower using the original (unapproved) draft ramp plan. After the first 2 hours of the | |||
downpower, the control room operator questioned plant conditions that were inconsistent with | |||
the simulator scenario and contacted the reactor engineer. The reactor engineer provided a | |||
copy of the approved ramp plan. No reactivity manipulations outside of the approved plan had | |||
been made. Operators continued the downpower using the approved ramp plan. | |||
The licensee performed an apparent cause evaluation and determined that the shift foreman did | |||
not validate that the ramp plan in use was the same as the one that the operations manager had | |||
approved. Licensee corrective actions included revising existing procedures to require | |||
validation of the ramp plan by the shift foreman and shift manager during the reactivity briefing. | |||
This event is also discussed in Diablo Canyon Power PlantNRC Integrated Inspection Report | |||
05000275/2009005 and 05000323/2009005, dated February 3, 2010 (ADAMS Accession | |||
No. ML100341199). | |||
===Arkansas Nuclear One=== | |||
On April 25, 2010, following the completion of a refueling outage, Arkansas Nuclear One, Unit 1 was at approximately 20-percent reactor power determined by heat balance (approximately | |||
30-percent reactor power indicated on nuclear instrumentation (NI)) and holding to allow | |||
instrumentation and controls (I&C) technicians to calibrate the NI, which involves adjusting the | |||
gain on the NI excore detectors so that NI indicated reactor power level matches the reactor | |||
power determined by heat balance. To prevent the integrated control system (ICS) from | |||
automatically moving control rods in response to the changing input of NI reactor power level | |||
from the gain adjustment, the calibration procedure first directs a control room operator to place | |||
the control rod station in manual. The I&C technician who was implementing the procedure | |||
stated to a control room operator, We are ready to place ICS to manual. The control room operator responded, ICS is in manual. However, this exchange did not result in the operator | |||
placing the control rod station in manual and it remained in automatic. When I&C technicians | |||
subsequently adjusted the gain on the NIs, control rods automatically withdrew for | |||
approximately 38 seconds and resulted in an automatic reactor trip because of high reactor | |||
power (49.55 percent NI indicated reactor power) and high RCS pressure. The rapid event | |||
succession did not afford operators time to complete diagnosis of the rod withdrawal and initiate | |||
manual corrective action. | |||
The causes of the event included failure to follow the NI calibration procedure, miscommunication between the I&C technician and the reactor operator, failure to conduct a | |||
pre-job brief, and lack of supervisory oversight. Additional information is available in Arkansas | |||
Nuclear OneNRC Integrated Inspection Report 05000313/2010003 and 05000368/2010003, dated August 5, 2010 (ADAMS Accession No. ML102180209). | |||
==BACKGROUND== | ==BACKGROUND== | ||
The following are related NRC generic communications: | The following are related NRC generic communications: | ||
* NRC IN 92-39, | * NRC IN 92-39, Unplanned Return to Criticality during Reactor Shutdown, dated | ||
* NRC IN 96-69, | |||
* NRC IN 97-62, | May 13, 1992, discussed events involving unplanned returns to criticality caused by the | ||
cooldown of the reactor coolant system during reactor shutdowns (ADAMS Accession | |||
No. ML031200314). | |||
* NRC IN 96-69, Operator Actions Affecting Reactivity, dated December 20, 1996, highlighted several events in which poor command and control during reactivity | |||
evolutions have led to unanticipated and unintended plant conditions (ADAMS | |||
Accession No. ML031050475). | |||
* NRC IN 97-62, Unrecognized Reactivity Addition During Plant Shutdown, dated | |||
August 6, 1997, discussed an incident in which a reactor operator inserted control rods | |||
to bring the reactor subcritical and then promptly withdrew the rods in order to take the | |||
reactor to the critical (ADAMS Accession No. ML031050177). | |||
==DISCUSSION== | ==DISCUSSION== | ||
One of the most important responsibilities of an on-duty licensed reactor operator and senior reactor operator is reactivity management in order to maintain the reactor in the desired condition, consistent with plant technical specifications, by properly anticipating, controlling, and responding to changing plant | One of the most important responsibilities of an on-duty licensed reactor operator and senior | ||
reactor operator is reactivity management in order to maintain the reactor in the desired | |||
condition, consistent with plant technical specifications, by properly anticipating, controlling, and | |||
responding to changing plant parameters. Reactivity management involves establishing and | |||
implementing procedures for operators to use in determining the effects on reactivity of plant | |||
changes, and to operate the controls associated with plant equipment that could affect | |||
reactivity. Before conducting planned evolutions involving reactivity changes (e.g., power | |||
decreases and increases), although there is no specific NRC requirement to do so, many | |||
licensee reactor engineering staffs prepare a reactivity management plan that helps control | |||
room operators maintain the reactor in the desired condition by providing expected plant | |||
responses and expected alarms. Required training is expected to give licensed operators an | |||
understanding of facility operating characteristics during steady-state and transient conditions, including causes and effects of temperature, pressure, coolant chemistry, and load changes, as well as, operating limitations and their bases. Licensee post-transient reviews are important for | |||
determining the cause of transients or unexpected plant responses and for taking corrective | |||
actions, such as procedure changes and training, to prevent recurrence. | |||
During one of the events discussed above, after the reactor became subcritical through xenon | |||
buildup and a reactor coolant temperature increase, operators delayed inserting control rods for | |||
nearly 2 hours. NRC IN 92-39 discusses an event in which, after the operators brought the | |||
reactor subcritical by inserting control rods, an inadvertent unplanned return to criticality | |||
occurred because operators delayed actions to continue inserting control rods while changing | |||
shifts. Although not specifically required, licensees may consider revising procedures and | |||
training operators so that, after the reactor becomes subcritical, the operators will proceed | |||
without delay to insert control rods or add boron to ensure the reactor remains shut down. | |||
==CONTACT== | ==CONTACT== | ||
This IN requires no specific action or written | 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/ by TMcGinty for | |||
Thomas B. Blount, Acting Director | |||
Division of Policy and Rulemaking | |||
Office of Nuclear Reactor Regulation | |||
===Technical Contact:=== | ===Technical Contact:=== | ||
== | ===Geoffrey Miller=== | ||
817-860-8141 geoffrey.miller@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. | |||
ML101810282 OFFICE BC:DRP:R-IV Tech Editor IOLB:DIRS BC:IOLB:DIRS | |||
NAME GMiller CHsu LVick JMcHale | |||
DATE 12/1/10 (e-mail) 9/4/10 (e-mail) 12/8/10 (e-mail) 12/8/10 (e-mail) | |||
OFFICE D:DIRS BC:SRXB:DSS OGC(NLO) OEGIB/DRA/RES | |||
NAME FBrown AUlses DRoth LCriscione | |||
Non-Concur | |||
DATE 1/20/2011 12/13/10 12/14/10 (e-mail) 12/15/10 | |||
OFFICE LA:PGCB:NRR PM:PGCB:NRR BC:PGCB:NRR D:DPR:NRR | |||
NAME CHawes DBeaulieu SRosenberg TBlount (TMcGinty for) | |||
OFFICE 01/25/2011 (e-mail) 1/20/2011 1/26/2011 1/31/2011}} | |||
{{Information notice-Nav}} | {{Information notice-Nav}} |
Latest revision as of 17:35, 13 November 2019
ML101810282 | |
Person / Time | |
---|---|
Issue date: | 01/31/2011 |
From: | Blount T Division of Policy and Rulemaking |
To: | |
Beaulieu, D P, NRR/DPR, 415-3243 | |
References | |
IN-11-002 | |
Download: ML101810282 (6) | |
UNITED STATES
NUCLEAR REGULATORY COMMISSION
OFFICE OF NUCLEAR REACTOR REGULATION
WASHINGTON, DC 20555-001 January 31, 2011 NRC INFORMATION NOTICE 2011-02: OPERATOR PERFORMANCE ISSUES
INVOLVING REACTIVITY MANAGEMENT AT
NUCLEAR POWER PLANTS
ADDRESSEES
All holders of operating licenses 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 events involving deficiencies with reactivity management planning and
implementation. 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.
DESCRIPTION OF CIRCUMSTANCES
Callaway Plant
During a Callaway Plant shutdown in October 2003, the control room operators did not
effectively control core reactivity during low-power operations. The event began on the morning
of October 20, 2003, when the Callaway Plant experienced an inverter failure on a
safety-related bus that put the unit in a 24-hour technical specification action to restore the
inverter or be in Mode 3 (hot standby) within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. The next morning, because the
inverter had not yet been restored, operators initiated a plant shutdown at approximately
10 percent per hour. With the main turbine on line and with turbine bypass valves closed, operators attempted to stabilize the plant at approximately 8-percent power. Per procedure, operators opened the turbine drain valves, which increased main steam flow, reducing reactor
coolant temperature and adding positive reactivity. At the same time, negative reactivity was
being inserted by xenon buildup, decreasing reactor power. The net effect was that reactor
coolant temperature decreased by approximately 10 degrees Fahrenheit over a half-hour
period. Operators did not withdraw control rods or dilute boron concentration to stabilize reactor
power or reactor coolant temperature. As a result of the lowering reactor coolant temperature, pressurizer level lowered enough to cause letdown to isolate. In addition, reactor coolant
temperature decreased below the technical specification required minimum temperature while
critical for several minutes. In response to the lowering reactor coolant temperature, the
operators manually tripped the main turbine with power at approximately 5 percent. Upon
tripping the main turbine, the turbine bypass valves controlled steam pressure, causing reactor
coolant temperature and pressurizer level to return to normal. Tripping the main turbine
reduced main steam flow, increasing reactor coolant temperature and adding negative
reactivity, which together with the addition of negative reactivity by xenon buildup, caused the
reactor to become subcritical. Operators did not insert the control rods until almost 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> after
the reactor became subcritical.
A subsequent review of this plant shutdown found that control room operators did not effectively
control reactivity to maintain the reactor in the desired condition during low-power operations by
properly anticipating, controlling, and responding to changing plant parameters. Operators did
not use control rods or boron concentrationtwo means that operators can use to directly
control the amount and timing of reactivity changesto adjust for reactivity changes by xenon
buildup and reactor coolant temperature changes. Specifically, reactivity was not effectively
controlled in that (1) operators did not sufficiently anticipate and compensate for xenon buildup
when they attempted to stabilize and hold the plant at approximately 8 percent power which
caused reactor coolant temperature to continue to decrease below the technical specification
required minimum temperature for criticality; (2) operators did not shut down the reactor in a
deliberate manner (e.g., by inserting control rod banks), but rather the reactor became
subcritical by xenon buildup and by the increase in reactor coolant temperature resulting from
the operators manually tripping the main turbine; and, (3) operators did not insert control rods
for nearly 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> after the reactor became subcritical. Fully inserting control rods provides
assurance that the reactor remains shut down (regardless of reactor coolant temperature or
xenon concentration). Notwithstanding, the NRC determined the delay in completing the
shutdown by inserting the control rods did not result in an unsafe reactor condition.
Operator performance in not effectively controlling reactivity was attributable, in part, to
weaknesses with management oversight, training, and procedural guidance. The pre-evolution
practice session using the simulator did not cover plant operations below 10-percent power and
did not include operation after the point where the operators tripped the main turbine. The
reactivity management plan did not address the possibility that the expected reactivity change
from tripping the main turbine together with the xenon buildup could cause the reactor to
become subcritical. The licensees initial post-shutdown review did not identify and evaluate the
atypical manner that the reactor became subcritical. This omission delayed application of the
lessons learned to operator qualification and requalification training and significantly delayed
procedure changes to address weaknesses in operator control of reactivity during low-power
operation. This event is also discussed in Callaway PlantNRC Integrated Inspection Report 05000483/2007003, dated August 2, 2007, which can be found on the NRCs public Web site in
the Agencywide Documents Access and Management System (ADAMS) under Accession
No. ML072140876.
River Bend Station
On March 8, 2008, with River Bend Station at 25-percent power, control room operators were
withdrawing control rods to increase reactor power. The operating procedure for plant startup
directs operators to withdraw control rods using a withdrawal sequence specified in a reactivity
control plan that is provided to them by licensee reactor engineering. However, the dedicated
reactor operator at the controls stated an incorrect target position when reading aloud a rod movement step in the reactivity control plan. As a result, this operator individually withdrew six
consecutive rods to position 24 rather than the target position 20 specified in the reactivity
control plan. The dedicated peer-check reactor operator did not identify that the stated target
position was incorrect because he could not readily see the reactivity control plan that was
resting on the lap of the reactor operator at the controls. The operator at the controls halted the
withdrawal of the seventh rod at position 18 after the dedicated peer-check reactor operator
identified the error. The licensee determined that the reactor operator at the controls and the
peer-checker did not follow the procedures to prevent human performance errors and that the
senior reactor operator did not maintain effective oversight of the activity. Additional information
is available in River Bend StationNRC Integrated Inspection Report 05000458/2008002, dated May 9, 2008 (ADAMS Accession No. ML081300838).
Diablo Canyon Power Plant, Unit 2
In August 2009, Diablo Canyon Power Plant Unit 2 was shut down in order to troubleshoot and
repair a main transformer bushing. In preparation for the shutdown, the control room operators
performed simulator training on a ramp downpower using a draft copy of a ramp plan provided
via e-mail by reactor engineering. Before the actual shutdown, a revised ramp plan was
provided by reactor engineering, approved by the operations manager, and issued in the shift
orders. This revised ramp plan was also e-mailed to all shift members. The oncoming shift
foreman and shift manager did not review the approved ramp plan located in the shift orders nor
did they review the ramp plan as part of the reactivity brief. Operators began the ramp
downpower using the original (unapproved) draft ramp plan. After the first 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> of the
downpower, the control room operator questioned plant conditions that were inconsistent with
the simulator scenario and contacted the reactor engineer. The reactor engineer provided a
copy of the approved ramp plan. No reactivity manipulations outside of the approved plan had
been made. Operators continued the downpower using the approved ramp plan.
The licensee performed an apparent cause evaluation and determined that the shift foreman did
not validate that the ramp plan in use was the same as the one that the operations manager had
approved. Licensee corrective actions included revising existing procedures to require
validation of the ramp plan by the shift foreman and shift manager during the reactivity briefing.
This event is also discussed in Diablo Canyon Power PlantNRC Integrated Inspection Report 05000275/2009005 and 05000323/2009005, dated February 3, 2010 (ADAMS Accession
No. ML100341199).
Arkansas Nuclear One
On April 25, 2010, following the completion of a refueling outage, Arkansas Nuclear One, Unit 1 was at approximately 20-percent reactor power determined by heat balance (approximately
30-percent reactor power indicated on nuclear instrumentation (NI)) and holding to allow
instrumentation and controls (I&C) technicians to calibrate the NI, which involves adjusting the
gain on the NI excore detectors so that NI indicated reactor power level matches the reactor
power determined by heat balance. To prevent the integrated control system (ICS) from
automatically moving control rods in response to the changing input of NI reactor power level
from the gain adjustment, the calibration procedure first directs a control room operator to place
the control rod station in manual. The I&C technician who was implementing the procedure
stated to a control room operator, We are ready to place ICS to manual. The control room operator responded, ICS is in manual. However, this exchange did not result in the operator
placing the control rod station in manual and it remained in automatic. When I&C technicians
subsequently adjusted the gain on the NIs, control rods automatically withdrew for
approximately 38 seconds and resulted in an automatic reactor trip because of high reactor
power (49.55 percent NI indicated reactor power) and high RCS pressure. The rapid event
succession did not afford operators time to complete diagnosis of the rod withdrawal and initiate
manual corrective action.
The causes of the event included failure to follow the NI calibration procedure, miscommunication between the I&C technician and the reactor operator, failure to conduct a
pre-job brief, and lack of supervisory oversight. Additional information is available in Arkansas
Nuclear OneNRC Integrated Inspection Report 05000313/2010003 and 05000368/2010003, dated August 5, 2010 (ADAMS Accession No. ML102180209).
BACKGROUND
The following are related NRC generic communications:
- NRC IN 92-39, Unplanned Return to Criticality during Reactor Shutdown, dated
May 13, 1992, discussed events involving unplanned returns to criticality caused by the
cooldown of the reactor coolant system during reactor shutdowns (ADAMS Accession
No. ML031200314).
- NRC IN 96-69, Operator Actions Affecting Reactivity, dated December 20, 1996, highlighted several events in which poor command and control during reactivity
evolutions have led to unanticipated and unintended plant conditions (ADAMS
Accession No. ML031050475).
- NRC IN 97-62, Unrecognized Reactivity Addition During Plant Shutdown, dated
August 6, 1997, discussed an incident in which a reactor operator inserted control rods
to bring the reactor subcritical and then promptly withdrew the rods in order to take the
reactor to the critical (ADAMS Accession No. ML031050177).
DISCUSSION
One of the most important responsibilities of an on-duty licensed reactor operator and senior
reactor operator is reactivity management in order to maintain the reactor in the desired
condition, consistent with plant technical specifications, by properly anticipating, controlling, and
responding to changing plant parameters. Reactivity management involves establishing and
implementing procedures for operators to use in determining the effects on reactivity of plant
changes, and to operate the controls associated with plant equipment that could affect
reactivity. Before conducting planned evolutions involving reactivity changes (e.g., power
decreases and increases), although there is no specific NRC requirement to do so, many
licensee reactor engineering staffs prepare a reactivity management plan that helps control
room operators maintain the reactor in the desired condition by providing expected plant
responses and expected alarms. Required training is expected to give licensed operators an
understanding of facility operating characteristics during steady-state and transient conditions, including causes and effects of temperature, pressure, coolant chemistry, and load changes, as well as, operating limitations and their bases. Licensee post-transient reviews are important for
determining the cause of transients or unexpected plant responses and for taking corrective
actions, such as procedure changes and training, to prevent recurrence.
During one of the events discussed above, after the reactor became subcritical through xenon
buildup and a reactor coolant temperature increase, operators delayed inserting control rods for
nearly 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />. NRC IN 92-39 discusses an event in which, after the operators brought the
reactor subcritical by inserting control rods, an inadvertent unplanned return to criticality
occurred because operators delayed actions to continue inserting control rods while changing
shifts. Although not specifically required, licensees may consider revising procedures and
training operators so that, after the reactor becomes subcritical, the operators will proceed
without delay to insert control rods or add boron to ensure the reactor remains shut down.
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/ by TMcGinty for
Thomas B. Blount, Acting Director
Division of Policy and Rulemaking
Office of Nuclear Reactor Regulation
Technical Contact:
Geoffrey Miller
817-860-8141 geoffrey.miller@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.
ML101810282 OFFICE BC:DRP:R-IV Tech Editor IOLB:DIRS BC:IOLB:DIRS
NAME GMiller CHsu LVick JMcHale
DATE 12/1/10 (e-mail) 9/4/10 (e-mail) 12/8/10 (e-mail) 12/8/10 (e-mail)
OFFICE D:DIRS BC:SRXB:DSS OGC(NLO) OEGIB/DRA/RES
NAME FBrown AUlses DRoth LCriscione
Non-Concur
DATE 1/20/2011 12/13/10 12/14/10 (e-mail) 12/15/10
OFFICE LA:PGCB:NRR PM:PGCB:NRR BC:PGCB:NRR D:DPR:NRR
NAME CHawes DBeaulieu SRosenberg TBlount (TMcGinty for)
OFFICE 01/25/2011 (e-mail) 1/20/2011 1/26/2011 1/31/2011