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

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)

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

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). 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

Division of Construction Inspection

and Operational Programs

Office of New Reactors

Lawrence E. Kokajko, Director

Division of Policy and Rulemaking

Office of Nuclear Reactor Regulation

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