Information Notice 2013-17, Significant Plant Transient Induced by Safety-Related Direct Current Bus Maintenance at Power: Difference between revisions

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{{#Wiki_filter:ML13193A009 UNITED STATES NUCLEAR REGULATORY COMMISSION OFFICE OF NUCLEAR REACTOR REGULATION OFFICE OF NEW REACTORS WASHINGTON, DC 20555-0001 September 6, 2013 NRC INFORMATION NOTICE 2013-17: SIGNIFICANT PLANT TRANSIENT INDUCED BY SAFETY-RELATED DIRECT CURRENT BUS MAINTENANCE AT POWER
{{#Wiki_filter:UNITED STATES
 
NUCLEAR REGULATORY COMMISSION
 
OFFICE OF NUCLEAR REACTOR REGULATION
 
OFFICE OF NEW REACTORS
 
WASHINGTON, DC 20555-0001 September 6, 2013 NRC INFORMATION NOTICE 2013-17:               SIGNIFICANT PLANT TRANSIENT INDUCED BY
 
SAFETY-RELATED DIRECT CURRENT BUS
 
MAINTENANCE AT POWER


==ADDRESSEES==
==ADDRESSEES==
All holders of an operating license or construction permit for a nuclear power reactor under Title 10 of the
All holders of an operating license or construction permit for a nuclear power reactor under


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 vessel.
Title 10 of the Code of Federal Regulations (10 CFR) Part 50, Domestic Licensing of


All holders of and applicants for a power reactor early site permit, combined license, standard design certification, standard design approval, or manufacturing license under 10 CFR Part 52,  
Production and Utilization Facilities, except those who have permanently ceased operations
"Licenses, Certifications, and Approvals for Nuclear Power Plants."
 
and have certified that fuel has been permanently removed from the reactor vessel.
 
All holders of and applicants for a power reactor early site permit, combined license, standard
 
design certification, standard design approval, or manufacturing license under 10 CFR Part 52, Licenses, Certifications, and Approvals for Nuclear Power Plants.


==PURPOSE==
==PURPOSE==
The U.S. Nuclear Regulatory Commission (NRC) is is
The U.S. Nuclear Regulatory Commission (NRC) is issuing this information notice (IN) to inform


suing this information notice (IN) to inform addressees of recent operating experience involving the loss of one train of a direct current
addressees of recent operating experience involving the loss of one train of a direct current


(DC) distribution system at power in a nuclear power plant. The NRC expects that recipients
(DC) distribution system at power in a nuclear power plant. The NRC expects that recipients


will review the information for applicability to their facilities and 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.
will review the information for applicability to their facilities and 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==


===Palisades Nuclear Plant===
===Palisades Nuclear Plant===
  Palisades Nuclear Plant was operating in Mode 1 at approximately 98 percent power on September 25, 2011. An electrician was performing maintenance on the energized left train
Palisades Nuclear Plant was operating in Mode 1 at approximately 98 percent power on
 
September 25, 2011. An electrician was performing maintenance on the energized left train


125-volt DC distribution Panel D11-2 (reference figure on next page). While in the process of
125-volt DC distribution Panel D11-2 (reference figure on next page). While in the process of


removing a breaker bus connector in safety-related Panel D11-2, without proper insulation of
removing a breaker bus connector in safety-related Panel D11-2, without proper insulation of


adjacent bus connectors, a maintenance technician inadvertently lost control of the energized positive copper bus connector, which swung down and contacted the negative copper bus connector. This resulted in an electrical fault in the panel and upstream shunt trip Breaker
adjacent bus connectors, a maintenance technician inadvertently lost control of the energized
 
positive copper bus connector, which swung down and contacted the negative copper bus
 
connector. This resulted in an electrical fault in the panel and upstream shunt trip Breaker


72-01 (between the DC bus and the battery) opened on over-current, disconnecting the battery
72-01 (between the DC bus and the battery) opened on over-current, disconnecting the battery


from the DC bus. By design, the fuse to Panel D11-2 was expected to isolate Panel D11-2 from the entire DC bus upon a fault in Panel D11-2. However, the shunt trip breaker was incorrectly installed in 1981 with thermal and instantaneous over-current protection features (latent design issue) and the over-current breaker trip device was at a current setting lower than the fuse
from the DC bus. By design, the fuse to Panel D11-2 was expected to isolate Panel D11-2 from
 
the entire DC bus upon a fault in Panel D11-2. However, the shunt trip breaker was incorrectly
 
installed in 1981 with thermal and instantaneous over-current protection features (latent design
 
issue) and the over-current breaker trip device was at a current setting lower than the fuse
 
current rating. Therefore, the shunt trip breaker opened quicker than the fuse could isolate the


current rating. Therefore, the shunt trip breaker opened quicker than the fuse could isolate the
fault in Panel D11-2 and the entire left train 125-volt DC bus was lost (reference figure below).


fault in Panel D11-2 and the entire left train 125-volt DC bus was lost (reference figure below). The electrical transient resulted in the failure of the in-service left train battery charger and associated left train inverters. This led to the loss of two preferred 120-volt alternating current
ML13193A009 The electrical transient resulted in the failure of the in-service left train battery charger and
 
associated left train inverters. This led to the loss of two preferred 120-volt alternating current


(AC) power sources (busses Y-10 and Y-30) that supply power for the left train of annunciation
(AC) power sources (busses Y-10 and Y-30) that supply power for the left train of annunciation


and instrumentation in the control room. The failure resulted in the operators having only the
and instrumentation in the control room. The failure resulted in the operators having only the


right train of equipment available for annunciation and instrumentation in the control room.
right train of equipment available for annunciation and instrumentation in the control room.


Left Train     Right Train
Left Train                                     Right Train


The loss of the left train 125-volt DC bus and two preferred AC power sources resulted in an
The loss of the left train 125-volt DC bus and two preferred AC power sources resulted in an


instantaneous reactor and turbine trip caused by a reactor protection system actuation. These
instantaneous reactor and turbine trip caused by a reactor protection system actuation. These


trips were coincident with the automatic actuation of the following systems and components: a
trips were coincident with the automatic actuation of the following systems and components: a


safety injection actuation signal (automatically started right train emergency core cooling systems); main steam isolation signal (closed ma
safety injection actuation signal (automatically started right train emergency core cooling


in steam isolation valves and main feedwater regulating valves); containment high radiation signal (switched control room heating, ventilation, and air conditioning system to emergency mode); containment isolation signal (closed left channel containment isolation valves); auxiliary feedwater (AFW) actuation signal (started one
systems); main steam isolation signal (closed main steam isolation valves and main feedwater


motor driven AFW pump and the turbine driven AFW pump with full flow); and a containment high pressure alarm (no actuation signal).  The protection circuitry actuated as a result of the loss of the left train of DC power and was not required to mitigate a degraded or abnormal condition of the reactor.  Also, at this time, the following major equipment was affected by the loss of the left train 125-volt DC system:  the primary coolant pumps 'A' and 'C' coasted down; nonsafety-related AC busses 1A and 1F did not fast transfer; and the atmospheric steam dump
regulating valves); containment high radiation signal (switched control room heating, ventilation, and air conditioning system to emergency mode); containment isolation signal (closed left


valve master controller lost power, which made them unavailable to provide a steaming path to reduce primary side temperature and pressure. Secondary side steam pressure was controlled
channel containment isolation valves); auxiliary feedwater (AFW) actuation signal (started one
 
motor driven AFW pump and the turbine driven AFW pump with full flow); and a containment
 
high pressure alarm (no actuation signal). The protection circuitry actuated as a result of the
 
loss of the left train of DC power and was not required to mitigate a degraded or abnormal condition of the reactor. Also, at this time, the following major equipment was affected by the
 
loss of the left train 125-volt DC system: the primary coolant pumps A and C coasted down;
nonsafety-related AC busses 1A and 1F did not fast transfer; and the atmospheric steam dump
 
valve master controller lost power, which made them unavailable to provide a steaming path to
 
reduce primary side temperature and pressure. Secondary side steam pressure was controlled


by the secondary side code safety valves for the first hour of the event.
by the secondary side code safety valves for the first hour of the event.
Line 78: Line 125:
During the transient, the operators encountered additional complications that included the
During the transient, the operators encountered additional complications that included the


following: an increasing primary coolant system (PCS) leak rate that was later determined to be
following: an increasing primary coolant system (PCS) leak rate that was later determined to be
 
from the actuation of a chemical and volume control system relief valve that relieved to the


from the actuation of a chemical and volume c
quench tank; an increasing PCS level in the pressurizer that could have reached a solid


ontrol system relief valve that relieved to the quench tank; an increasing PCS level in the pressurizer that could have reached a solid condition; an increasing steam generator 'A' level, which reached approximately 95 percent; and the actuation of suction and discharge pressure relief valves for the charging pumps, which
condition; an increasing steam generator A level, which reached approximately 95 percent; and
 
the actuation of suction and discharge pressure relief valves for the charging pumps, which


displaced volume control tank water into the charging pump cubicles located in the auxiliary
displaced volume control tank water into the charging pump cubicles located in the auxiliary
Line 88: Line 139:
building.
building.


The operators immediately responded to the event by using the emergency operating procedures. However, the licensee's functional recovery and associated off-normal procedures
The operators immediately responded to the event by using the emergency operating
 
procedures. However, the licensees functional recovery and associated off-normal procedures


were only written to address the loss of a DC train coincident with the loss of only one preferred
were only written to address the loss of a DC train coincident with the loss of only one preferred


AC power source. In this case two preferred AC power sources were lost, thus complicating the
AC power source. In this case two preferred AC power sources were lost, thus complicating the
 
operators response. Operators were able to implement the existing procedures, while
 
maintenance personnel troubleshot, diagnosed and corrected the loss of the 125-volt DC


operators' response. Operators were able to implement the existing procedures, while
system. This action took approximately 50 minutes. Once power was restored to the preferred


maintenance personnel troubleshot, diagnosed and corrected the loss of the 125-volt DC system.  This action took approximately 50 minutes.  Once power was restored to the preferred AC busses, secondary-side steam pressure was controlled through the use of the atmospheric
AC busses, secondary-side steam pressure was controlled through the use of the atmospheric


steam dump valves. The operators then safely brought the plant to cold shutdown and the
steam dump valves. The operators then safely brought the plant to cold shutdown and the


licensee began an event investigation. On Sept
licensee began an event investigation. On September 26, 2011, the NRC dispatched a special


ember 26, 2011, the NRC dispatched a special inspection team to independently review the licensee's actions and to determine the causes of
inspection team to independently review the licensees actions and to determine the causes of


the event.
the event.


Additional information appears in the following NRC inspection reports:  
Additional information appears in the following NRC inspection reports:
Palisades Nuclear Plant - NRC Special Inspection Team (SIT) Report 05000255/2011014 Preliminary Yellow Finding, dated November 29, 2011, in the NRCs Agencywide Documents
 
Access and Management System (ADAMS) under Accession No. ML113330802.


"Palisades Nuclear Plant - NRC Special Inspection Team (SIT) Report 05000255/2011014 Preliminary Yellow Finding," dated November 29, 2011, in the NRC's Agencywide Documents Access and Management System (ADAMS) under Accession No. ML113330802.
Final Significance Determination of Yellow and White Findings with Assessment Followup and


"Final Significance Determination of Yellow and White Findings with Assessment Followup and
Notice of Violation, NRC Inspection Report Nos. 05000255/2011019 and 05000255/2011020,
Palisades Nuclear Plant, dated February 14, 2012, under ADAMS Accession


===Notice of Violation, NRC Inspec===
No. ML120450037.
tion Report Nos. 05000255/2011019 and 05000255/2011020, Palisades Nuclear Plant," dated February 14, 2012, under ADAMS Accession No. ML120450037.


==BACKGROUND==
==BACKGROUND==
The DC power system provides power for Class 1E equipment such as breaker control, the
The DC power system provides power for Class 1E equipment such as breaker control, the


plant instrumentation and control, monitoring, lighting (main control room and remote shutdown area) and other functions.  The battery supplies the load without interruption should the battery charger or associated preferred AC source fail.
plant instrumentation and control, monitoring, lighting (main control room and remote shutdown


Criterion 21, "Protection System Reliability and Testability," of Appendix A, "General Design
area) and other functions. The battery supplies the load without interruption should the battery


Criteria for Nuclear Power Plants," to 10 CFR Part 50 states, "The protection system shall be designed for high functional reliability and inserv
charger or associated preferred AC source fail.


ice testability commensurate with the safety functions to be performed."
Criterion 21, Protection System Reliability and Testability, of Appendix A, General Design
 
Criteria for Nuclear Power Plants, to 10 CFR Part 50 states, The protection system shall be
 
designed for high functional reliability and inservice testability commensurate with the safety
 
functions to be performed.


==DISCUSSION==
==DISCUSSION==
Reliability of the Class 1E DC power system is important in a nuclear power generating station. The DC power system is designed so that no single failure of an electrical panel, battery, or a
Reliability of the Class 1E DC power system is important in a nuclear power generating station.
 
The DC power system is designed so that no single failure of an electrical panel, battery, or a


battery charger will result in a condition that will prevent the safe shutdown of the plant.
battery charger will result in a condition that will prevent the safe shutdown of the plant.


==Addressees==
==Addressees==
are encouraged to review their design, installed equipment, processes, and procedures to ensure that a similar latent design issue would not degrade the performance of
are encouraged to review their design, installed equipment, processes, and
 
procedures to ensure that a similar latent design issue would not degrade the performance of


their DC electrical system.
their DC electrical system.


While the cause of the event at the Palisades Nuclear Plant revealed a latent design deficiency with DC protective devices, licensee current performance issues directly initiated and contributed to the event. The work performed could have been completed safely and without
While the cause of the event at the Palisades Nuclear Plant revealed a latent design deficiency
 
with DC protective devices, licensee current performance issues directly initiated and
 
contributed to the event. The work performed could have been completed safely and without


incident had licensee personnel followed and implemented existing procedures, as required.
incident had licensee personnel followed and implemented existing procedures, as required.


Several operating experience lessons learned for emergent safety-related work on plant equipment were identified as a result of this plant transient. The following are applicable to all
Several operating experience lessons learned for emergent safety-related work on plant
 
equipment were identified as a result of this plant transient. The following are applicable to all
 
NRC licensees:
    *  In-field, engaged management oversight that reinforces worker behaviors and approved
 
procedures is critical when working safely on risk significant, safety-related equipment.
 
*  Ensure operations procedures exist for the response to all design basis events and
 
licensed operators are trained on those scenarios as part of initial or continuing training.
 
*  All employees are essential and responsible for ensuring work control processes and


NRC licensees:
procedures are followed for all aspects of a maintenance evolution including work order


* In-field, engaged management oversight that reinforces worker behaviors and approved procedures is critical when working safely on risk significant, safety-related equipment.
planning, pre-job brief level determination, operational risk assessment of the work to be


* Ensure operations procedures exist for the response to all design basis events and licensed operators are trained on those scenarios as part of initial or continuing training.
performed, conducting pre-job briefs and in-field changes to work orders or procedures.


* All employees are essential and responsible for ensuring work control processes and procedures are followed for all aspects of a maintenance evolution including work order planning, pre-job brief level determination, operational risk assessment of the work to be performed, conducting pre-job briefs and in-field changes to work orders or procedures.
*   Noncovered nuclear workers (i.e., not covered under 10 CFR Part 26, Fitness for Duty


* Noncovered nuclear workers (i.e., not covered under 10 CFR Part 26, "Fitness for Duty Programs"), including department-level managers and above, are required to follow applicable work hour guidance for noncovered workers as prescribed in licensee
Programs), including department-level managers and above, are required to follow
 
applicable work hour guidance for noncovered workers as prescribed in licensee


procedures.
procedures.


* Before performing maintenance activities, an effective operational risk assessment typically includes worse-case scenarios, the importance of the work, and specified measures for working safely on risk significant equipment.
*   Before performing maintenance activities, an effective operational risk assessment
 
typically includes worse-case scenarios, the importance of the work, and specified
 
measures for working safely on risk significant equipment.


==CONTACT==
==CONTACT==
Line 161: Line 254:
matter to the technical contact listed below or to the appropriate NRC project manager.
matter to the technical contact listed below or to the appropriate NRC project manager.


/RA/       /RA/  
/RA/                                                 /RA/
Michael C. Cheok, Acting Director
Michael C. Cheok, Acting Director                    Lawrence E. Kokajko, Director


Division of Construction Inspection
Division of Construction Inspection                   Division of Policy and Rulemaking


and Operational Programs
and Operational Programs                             Office of Nuclear Reactor Regulation


Office of New Reactors
===Office of New Reactors===


Lawrence E. Kokajko, Director
===Technical Contact:===


Division of Policy and Rulemaking Office of Nuclear Reactor Regulation
===Robert G. Krsek, Region III===
                    920-388-3156 Robert.Krsek@nrc.gov


===Technical Contact:===
Note: NRC generic communications may be found on the NRC public Web site, http://www.nrc.gov, under NRC Library/Document Collections.
Robert G. Krsek, Region III        920-388-3156 Robert.Krsek@nrc.gov
 
ML13193A009            *concurred via e-mail      TAC No. MF2430
OFFICE      RIII/DRP/B2    NRR/DRA/AHPB          Tech Editor    RIII/DRP/B4/BC DE/EEEB/(A)BC
 
NAME          RKrsek*          MKeefe*          RBrennan*        JGiessner*      RMathew
 
DATE          08/16/13          08/21/13          08/01/13          08/19/13      07/24/13 OFFICE DRA/AHPB/BC            NRR/DRA/D        NRR/PGCB/LA NRR/PGCB/PM NRR/PGCB/(A)BC
 
NAME          UShoop*      JGiitter (SLee for)    CHawes            TAlexion      SStuchell
 
DATE          08/20/13          08/27/13          08/28/13          08/28/13      08/29/13 OFFICE NRO/DCIP/(A)D        NRR/DPR/DD        NRR/DPR/D


Note:  NRC generic communications may be found on the NRC public Web site, http://www.nrc.gov, under NRC Library/Document Collections.
NAME          MCheok          SBahadur          LKokajko


ML13193A009      *concurred via e-mail        TAC No. MF2430 OFFICE RIII/DRP/B2 NRR/DRA/AHPBTech Editor RIII/DRP/B4/BC DE/EEEB/(A)BCNAME RKrsek* MKeefe* RBrennan* JGiessner* RMathew DATE 08/16/13 08/21/13 08/01/13 08/19/13 07/24/13 OFFICE DRA/AHPB/BC NRR/DRA/D NRR/PGCB/LANRR/PGCB/PM NRR/PGCB/(A)BCNAME UShoop* JGiitter (SLee for)CHawes TAlexion SStuchell DATE 08/20/13 08/27/13 08/28/13 08/28/13 08/29/13 OFFICE NRO/DCIP/(A)D NRR/DPR/DD NRR/DPR/D  NAME MCheok SBahadur LKokajko  DATE 09/03/13 09/04/13 09/ 06 /13}}
DATE         09/03/13         09/04/13         09/ 06 /13}}


{{Information notice-Nav}}
{{Information notice-Nav}}

Revision as of 17:05, 4 November 2019

Significant Plant Transient Induced by Safety-Related Direct Current Bus Maintenance at Power
ML13193A009
Person / Time
Issue date: 09/06/2013
From: Michael Cheok, Kokajko L
NRC/NRR/DCI, Generic Communications Projects Branch
To:
Alexion T
References
IN-13-017
Download: ML13193A009 (8)
v • d • e


UNITED STATES

NUCLEAR REGULATORY COMMISSION

OFFICE OF NUCLEAR REACTOR REGULATION

OFFICE OF NEW REACTORS

WASHINGTON, DC 20555-0001 September 6, 2013 NRC INFORMATION NOTICE 2013-17: SIGNIFICANT PLANT TRANSIENT INDUCED BY

SAFETY-RELATED DIRECT CURRENT BUS

MAINTENANCE AT POWER

ADDRESSEES

All holders of an operating license or construction permit for a nuclear power reactor under

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 vessel.

All holders of and applicants for a power reactor early site permit, combined license, standard

design certification, standard design approval, or manufacturing license under 10 CFR Part 52, Licenses, Certifications, and Approvals for Nuclear Power Plants.

PURPOSE

The U.S. Nuclear Regulatory Commission (NRC) is issuing this information notice (IN) to inform

addressees of recent operating experience involving the loss of one train of a direct current

(DC) distribution system at power in a nuclear power plant. The NRC expects that recipients

will review the information for applicability to their facilities and 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

Palisades Nuclear Plant

Palisades Nuclear Plant was operating in Mode 1 at approximately 98 percent power on

September 25, 2011. An electrician was performing maintenance on the energized left train

125-volt DC distribution Panel D11-2 (reference figure on next page). While in the process of

removing a breaker bus connector in safety-related Panel D11-2, without proper insulation of

adjacent bus connectors, a maintenance technician inadvertently lost control of the energized

positive copper bus connector, which swung down and contacted the negative copper bus

connector. This resulted in an electrical fault in the panel and upstream shunt trip Breaker

72-01 (between the DC bus and the battery) opened on over-current, disconnecting the battery

from the DC bus. By design, the fuse to Panel D11-2 was expected to isolate Panel D11-2 from

the entire DC bus upon a fault in Panel D11-2. However, the shunt trip breaker was incorrectly

installed in 1981 with thermal and instantaneous over-current protection features (latent design

issue) and the over-current breaker trip device was at a current setting lower than the fuse

current rating. Therefore, the shunt trip breaker opened quicker than the fuse could isolate the

fault in Panel D11-2 and the entire left train 125-volt DC bus was lost (reference figure below).

ML13193A009 The electrical transient resulted in the failure of the in-service left train battery charger and

associated left train inverters. This led to the loss of two preferred 120-volt alternating current

(AC) power sources (busses Y-10 and Y-30) that supply power for the left train of annunciation

and instrumentation in the control room. The failure resulted in the operators having only the

right train of equipment available for annunciation and instrumentation in the control room.

Left Train Right Train

The loss of the left train 125-volt DC bus and two preferred AC power sources resulted in an

instantaneous reactor and turbine trip caused by a reactor protection system actuation. These

trips were coincident with the automatic actuation of the following systems and components: a

safety injection actuation signal (automatically started right train emergency core cooling

systems); main steam isolation signal (closed main steam isolation valves and main feedwater

regulating valves); containment high radiation signal (switched control room heating, ventilation, and air conditioning system to emergency mode); containment isolation signal (closed left

channel containment isolation valves); auxiliary feedwater (AFW) actuation signal (started one

motor driven AFW pump and the turbine driven AFW pump with full flow); and a containment

high pressure alarm (no actuation signal). The protection circuitry actuated as a result of the

loss of the left train of DC power and was not required to mitigate a degraded or abnormal condition of the reactor. Also, at this time, the following major equipment was affected by the

loss of the left train 125-volt DC system: the primary coolant pumps A and C coasted down;

nonsafety-related AC busses 1A and 1F did not fast transfer; and the atmospheric steam dump

valve master controller lost power, which made them unavailable to provide a steaming path to

reduce primary side temperature and pressure. Secondary side steam pressure was controlled

by the secondary side code safety valves for the first hour of the event.

During the transient, the operators encountered additional complications that included the

following: an increasing primary coolant system (PCS) leak rate that was later determined to be

from the actuation of a chemical and volume control system relief valve that relieved to the

quench tank; an increasing PCS level in the pressurizer that could have reached a solid

condition; an increasing steam generator A level, which reached approximately 95 percent; and

the actuation of suction and discharge pressure relief valves for the charging pumps, which

displaced volume control tank water into the charging pump cubicles located in the auxiliary

building.

The operators immediately responded to the event by using the emergency operating

procedures. However, the licensees functional recovery and associated off-normal procedures

were only written to address the loss of a DC train coincident with the loss of only one preferred

AC power source. In this case two preferred AC power sources were lost, thus complicating the

operators response. Operators were able to implement the existing procedures, while

maintenance personnel troubleshot, diagnosed and corrected the loss of the 125-volt DC

system. This action took approximately 50 minutes. Once power was restored to the preferred

AC busses, secondary-side steam pressure was controlled through the use of the atmospheric

steam dump valves. The operators then safely brought the plant to cold shutdown and the

licensee began an event investigation. On September 26, 2011, the NRC dispatched a special

inspection team to independently review the licensees actions and to determine the causes of

the event.

Additional information appears in the following NRC inspection reports:

Palisades Nuclear Plant - NRC Special Inspection Team (SIT) Report 05000255/2011014 Preliminary Yellow Finding, dated November 29, 2011, in the NRCs Agencywide Documents

Access and Management System (ADAMS) under Accession No. ML113330802.

Final Significance Determination of Yellow and White Findings with Assessment Followup and

Notice of Violation, NRC Inspection Report Nos. 05000255/2011019 and 05000255/2011020,

Palisades Nuclear Plant, dated February 14, 2012, under ADAMS Accession

No. ML120450037.

BACKGROUND

The DC power system provides power for Class 1E equipment such as breaker control, the

plant instrumentation and control, monitoring, lighting (main control room and remote shutdown

area) and other functions. The battery supplies the load without interruption should the battery

charger or associated preferred AC source fail.

Criterion 21, Protection System Reliability and Testability, of Appendix A, General Design

Criteria for Nuclear Power Plants, to 10 CFR Part 50 states, The protection system shall be

designed for high functional reliability and inservice testability commensurate with the safety

functions to be performed.

DISCUSSION

Reliability of the Class 1E DC power system is important in a nuclear power generating station.

The DC power system is designed so that no single failure of an electrical panel, battery, or a

battery charger will result in a condition that will prevent the safe shutdown of the plant.

Addressees

are encouraged to review their design, installed equipment, processes, and

procedures to ensure that a similar latent design issue would not degrade the performance of

their DC electrical system.

While the cause of the event at the Palisades Nuclear Plant revealed a latent design deficiency

with DC protective devices, licensee current performance issues directly initiated and

contributed to the event. The work performed could have been completed safely and without

incident had licensee personnel followed and implemented existing procedures, as required.

Several operating experience lessons learned for emergent safety-related work on plant

equipment were identified as a result of this plant transient. The following are applicable to all

NRC licensees:

  • In-field, engaged management oversight that reinforces worker behaviors and approved

procedures is critical when working safely on risk significant, safety-related equipment.

  • Ensure operations procedures exist for the response to all design basis events and

licensed operators are trained on those scenarios as part of initial or continuing training.

  • All employees are essential and responsible for ensuring work control processes and

procedures are followed for all aspects of a maintenance evolution including work order

planning, pre-job brief level determination, operational risk assessment of the work to be

performed, conducting pre-job briefs and in-field changes to work orders or procedures.

  • Noncovered nuclear workers (i.e., not covered under 10 CFR Part 26, Fitness for Duty

Programs), including department-level managers and above, are required to follow

applicable work hour guidance for noncovered workers as prescribed in licensee

procedures.

  • Before performing maintenance activities, an effective operational risk assessment

typically includes worse-case scenarios, the importance of the work, and specified

measures for working safely on risk significant equipment.

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 NRC project manager.

/RA/ /RA/

Michael C. Cheok, Acting Director Lawrence E. Kokajko, Director

Division of Construction Inspection Division of Policy and Rulemaking

and Operational Programs Office of Nuclear Reactor Regulation

Office of New Reactors

Technical Contact:

Robert G. Krsek, Region III

920-388-3156 Robert.Krsek@nrc.gov

Note: NRC generic communications may be found on the NRC public Web site, http://www.nrc.gov, under NRC Library/Document Collections.

ML13193A009 *concurred via e-mail TAC No. MF2430

OFFICE RIII/DRP/B2 NRR/DRA/AHPB Tech Editor RIII/DRP/B4/BC DE/EEEB/(A)BC

NAME RKrsek* MKeefe* RBrennan* JGiessner* RMathew

DATE 08/16/13 08/21/13 08/01/13 08/19/13 07/24/13 OFFICE DRA/AHPB/BC NRR/DRA/D NRR/PGCB/LA NRR/PGCB/PM NRR/PGCB/(A)BC

NAME UShoop* JGiitter (SLee for) CHawes TAlexion SStuchell

DATE 08/20/13 08/27/13 08/28/13 08/28/13 08/29/13 OFFICE NRO/DCIP/(A)D NRR/DPR/DD NRR/DPR/D

NAME MCheok SBahadur LKokajko

DATE 09/03/13 09/04/13 09/ 06 /13


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