Information Notice 2018-03, Operating Experience Regarding Failure to Meet Technical Specifications Requirements for Changing Plant Conditions

Operating Experience Regarding Failure to Meet Technical Specifications Requirements for Changing Plant Conditions

ML17303A791
Person / Time
Issue date:	02/26/2018
From:	Mcginty T, Chris Miller Division of Construction Inspection and Operational Programs, Division of Inspection and Regional Support
To:
Tanya Mensah
References
IN-18-003
Download: ML17303A791 (5)

UNITED STATES

NUCLEAR REGULATORY COMMISSION

OFFICE OF NUCLEAR REACTOR REGULATION

OFFICE OF NEW REACTORS

WASHINGTON, DC 20555-0001 February 26, 2018 NRC INFORMATION NOTICE 2018-03: OPERATING EXPERIENCE REGARDING

FAILURE TO MEET TECHNICAL

SPECIFICATIONS REQUIREMENTS FOR

CHANGING PLANT CONDITIONS

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 that have permanently ceased operations

and have certified that fuel has been permanently removed from the reactor vessel.

All holders of an operating license for a nonpower reactor (research reactor, test reactor, or

critical assembly) under 10 CFR Part 50, except those that have permanently ceased

operations.

All holders of and applicants for a combined 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 several recent events in which operators failed to ensure that the requirements of

the plant technical specifications (TS) were met as the plant conditions changed. The NRC

expects recipients to review the information for applicability to their facilities and to consider

actions, as appropriate, to avoid similar problems. However, suggestions contained in this IN

are not NRC requirements; therefore, no specific action or written response is required.

DESCRIPTION OF CIRCUMSTANCES

Watts Bar Nuclear Plant, Unit 1

On October 22, 2015, the operating crew at the Watts Bar Nuclear Plant, Unit 1 (Watts Bar)

determined that both source range (SR) level trip channels (N-31 and N-32) were in the bypass

position with the reactor at 27-percent rated thermal power (RTP). The SR level trip switches

were left in bypass, outside of their required configuration, thereby removing a trip function

required by the TS during rod withdrawal. An incident prompt investigation led by Tennessee

Valley Authority (the licensee) determined that the SR trip functions were inoperable from the

time the reactor trip breakers (RTBs) were closed in Mode 3 on October 19, 2015, until reactor

power exceeded the P-6 permissive interlock (1.66 x 10-4 percent RTP) on October 21, 2015.

ML17303A791 TS 3.3.1, Reactor Trip System Instrumentation, requires these SR trip functions to be operable

with the RTBs closed in Modes 5, 4, 3, and 2 until reactor power exceeds the P-6 (the SR block

permissive) interlock. On October 21, 2015, at 0346 Eastern Daylight Time, the plant entered

Mode 2 with both SR level trip channels inoperable. This was a mode change violation in

accordance with limiting condition for operation (LCO) 3.0.4. Under these conditions, TS 3.3.1, Required Actions 4.1, 1.1 and J.1, required that the operators take immediate actions to

suspend operations involving positive reactivity additions and open the RTBs. Two lit

annunciators on the main control board indicated the bypass condition, but the operators did not

take immediate action to open the RTB because they did not notice that the SR level trip

channels were bypassed. In addition, operators failed to note the lit annunciators during shift

turnovers and board walkdowns from October 19 through October 22, 2015, and during

completion of the mode change checklist board walkdown on October 20, 2015, to transition

from Mode 3 to Mode 2. By the time the operators recognized the improper configuration, reactor power was above the P-6 permissive interlock, and the SR trip functions were no longer

a TS requirement.

The licensee identified that the operators failed to identify a bypassed safety function during

reactor testing and start-up due to inadequate tracking and validation of essential information.

Inadequate operating procedures used to control SR level trip switches prior to core reload were

identified as a contributing cause that led to this event. Specifically, on October 7, 2015, before

core reload, the SR level trips had been placed in bypass, in accordance with the licensees

procedure for power escalation testing to avoid spurious trips and alarms during the core reload

process. On October 9, 2015, core reload was completed, at which point the switches should

have been returned to the normal position. However, a revision to the procedure for power

escalation testing back in July 2013 had inadvertently omitted the step to return the switches to

normal. This issue was entered into the licensees corrective action program.

Further details appear in Watts Bar Licensee Event Report 05000390/2015-006-00, dated

December 21, 2015, and in NRC Integrated Inspection Report 05000390, 05000391/2015004, dated February 12, 2016. These documents are available on the NRCs public Web site under

Agencywide Documents Access and Management System (ADAMS) Accession

Nos. ML15355A525 and ML16043A214, respectively.

Davis-Besse Nuclear Power Station, Unit 1 On May 10, 2016, at approximately 0528 hours0.00611 days <br />0.147 hours <br />8.730159e-4 weeks <br />2.00904e-4 months <br />, the Davis-Besse Nuclear Power Station

(Davis-Besse) was at approximately 53 percent power and was increasing power following a

refueling outage. During a walkdown of control room indications, a Davis-Besse senior

manager determined that all four anticipatory reactor trip system (ARTS) instrumentation

channels were in bypass. The ARTS initiates a reactor trip following a turbine trip or loss of

both main feed pumps in order to reduce the magnitude of pressure and temperature transients

on the reactor coolant system and lower the probability of a pressurizer pilot-operated relief

valve actuation during these events.

The turbine trip function trips the reactor when the main turbine is lost at high power levels in

anticipation of the associated loss of heat sink. TS 3.3.16, Anticipatory Reactor Trip System

Instrumentation, requires the ARTS turbine trip function to be in normal when the plant is

operating above 45 percent power. As with the Watts Bar event discussed above, the ARTS

channels had been placed in bypass to support work during a refueling outage; however, the

procedure did not address the need to restore the system to a normal state upon completion of

the work. FirstEnergy Nuclear Operating Company (the licensee) investigated this event and determined

that multiple operators were aware that the ARTS was in bypass. Other operators were aware

that the annunciator alarm for the ARTS in bypass was lit and that it remained lit after the main

feed pump trip input for the ARTS was taken from bypass to normal during the startup. All of

the operators involved assumed that the startup procedure would direct the system to be placed

in normal. In fact, the applicable procedures required an operator to verify that the system was

returned to normal before the reactor reaches 40 percent power. However, the operator on shift

at that time misinterpreted the requirement as not applicable because the reactor had not yet

reached 40 percent power.

The licensee determined that the operators failed to work together effectively as a team to

ensure that the ARTS was operable before the plant entered the mode of applicability during

startup. The system was placed in normal upon discovery at 53 percent power to restore

operability. This issue was entered into the licensees corrective action program.

Further details on this event appear in Davis-Besse Licensee Event

Report 05000346/2016-005-00, dated July 11, 2016 (ADAMS Accession No. ML16194A343),

and in NRC Integrated Inspection Report 05000346/2016003, dated November 4, 2016 (ADAMS Accession No. ML16309A098).

Grand Gulf Nuclear Station, Unit 1

On September 8, 2016, Entergy Operations, Inc. (the licensee), manually shut down the Grand

Gulf Nuclear Station (Grand Gulf, Unit 1) reactor to replace pump A of the residual heat removal

(RHR) system after a failed TS surveillance. Grand Gulf TS 3.4.10, Residual Heat Removal

(RHR) Shutdown Cooling SystemCold Shutdown, requires two trains of RHR to be operable

when the reactor is shut down. With the A train of the RHR system inoperable for the pump

replacement, the TS requires verification that an alternate method of decay heat removal is

available. This action must be completed within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> of one RHR pump being taken out of

service and once every 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> thereafter while the plant is in Mode 4. To meet this

requirement, the licensee placed the alternate decay heat removal (ADHR) system in standby

after the plant reached Mode 4 on September 9, 2016, and verified its availability daily until the

A train of the RHR system was restored to operability on September 23, 2016.

Following replacement of the A RHR system pump and while attempting to place the ADHR

system in operation, licensee personnel identified that the ADHR system had not actually been

available as an alternate method of decay heat removal. Since August 10, 2016, the ADHR

heat exchanger tube-side cooling water system had been clearance tagged as closed to support

cleaning of the heat exchanger tubes. The licensees daily verification of the availability of the

ADHR system in accordance with the TS action statement had been administrative in nature.

The licensee did not require a physical walkdown of the ADHR system to verify its availability.

In addition, although the operators verified that no clearance tagouts were impacting the ADHR

system, they failed to consider that a clearance tagout on the plant service water system could

also affect the availability of the ADHR system. The licensee determined that when the work

requiring the tagout had been completed on August 10, 2016, the work complete box had

never been checked; therefore, the tags were left hanging. The licensees procedure for placing

the ADHR system in standby did not specify that the plant service water isolation valves for the

ADHR heat exchanger needed to be open. This issue was entered into the licensees corrective

action program. Additional information appears in the Grand Gulf Licensee Event Report

05000416/2016-008-01, dated August 16, 2017 (ADAMS Accession No. ML17228A233),

and in NRC Special Inspection Report 05000416/2016008, dated October 27, 2017 (ADAMS

Accession No. ML17303B200).

DISCUSSION

As specified in 10 CFR 50.36, Technical specifications, plant TS are derived from the analysis

and evaluation included in the plant safety analysis report and constitute a part of the license

authorizing operation of each reactor plant. Each LCO in the TS lists the mode(s) of

applicability for that condition and the actions to be taken if the conditions are not met. In each

event summarized above, the licensees took systems out of service for activities while the

plants were in a mode that did not require the systems. In doing so, the licensees failed to

ensure that the systems were restored to operable when required by TS LCOs.

Multiple administrative practices, including maintenance tracking systems, operating

procedures, operational checklists, and operator walkdowns, aid operators in maintaining an

understanding of the configuration of plant systems to ensure TS operability before entering a

mode of applicability that requires a system. In each case discussed in this IN, these practices

failed to accomplish their intended function when procedures were incomplete or misunderstood

and when plant personnel failed to effectively and aggressively communicate concerns about

unexpected tags, alarms, or indications. Industry operating experience has shown that best

practices, such as noting conditional LCOs when taking equipment out of service that is

required in another mode, caution-tagging equipment in an abnormal alignment, and issuing

return-to-service checklists to ensure that systems are returned to their expected condition

when work is complete, provide additional barriers to configuration control issues and

noncompliance with TS operability requirements. When followed appropriately, these practices

may significantly reduce the potential for TS violations.

CONTACT

This IN requires no specific action or written response. Please direct any questions about this

matter to the technical contacts listed below or the appropriate Office of Nuclear Reactor

Regulation (NRR) project manager.

/RA/ (Paul G. Krohn for) /RA/

Timothy J. McGinty, Director Christopher G. Miller, Director

Division of Construction Inspection Division of Inspection and Regional Support

and Operational Programs Office of Nuclear Reactor Regulation

Office of New Reactors

Technical Contacts: Rebecca Sigmon, NRR Margaret Chernoff, NRR

301-415-0895 301-415-2240

Rebecca.Sigmon@nrc.gov Margaret.Chernoff@nrc.gov

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

ML17303A791 *via email

OFFICE TECH EDITOR* NRR/DIRS/IOEB* NRR/DSS/SSTB* NRR/DSS/SSTB/BC* NRR/IOEB/DIRS/BC*

NAME JDougherty RSigmon MChernoff VCusumano RElliott

DATE 11/22/17 12/21/17 12/22/17 1/12/18 12/22/17 OFFICE NRR/DIRS/IRGB/PM NRR/DIRS/IRGB/LA NRR/DIRS/IRGB/BC NRO/DCIP/D NRR/DIRS/D

NAME TMensah* ELee HChernoff TMcGinty (PKroh for) CMiller

DATE 1/12/18 1/29/18 2/6/18 2/20/18 2/26/18