Information Notice 2012-21, Reactor Vessel Closure Head Studs Remain Detensioned During Plant Startup

Reactor Vessel Closure Head Studs Remain Detensioned During Plant Startup

ML12264A518
Person / Time
Issue date:	12/10/2012
From:	Laura Dudes, Mcginty T Division of Construction Inspection and Operational Programs, Division of Policy and Rulemaking
To:
Alexion, T
References
IN-12-20
Download: ML12264A518 (6)

ML12264A518 UNITED STATES

NUCLEAR REGULATORY COMMISSION

OFFICE OF NUCLEAR REACTOR REGULATION

OFFICE OF NEW REACTORS

WASHINGTON, DC 20555-0001

December 10, 2012

NRC INFORMATION NOTICE 2012-21:

REACTOR VESSEL CLOSURE HEAD STUDS

REMAIN DETENSIONED DURING PLANT

STARTUP

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 or applicants for an early site permit, standard design certification, standard

design approval, manufacturing license, or 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 an event involving detensioned reactor vessel closure head studs at a

boiling-water reactor that resulted in leakage from the reactor vessel during startup operations

and a manual scram. 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

Brunswick Steam Electric Plant, Unit 2

On November 16, 2011, the Brunswick Steam Electric Plant, Unit 2 (Brunswick 2, a General

Electric boiling-water reactor) was in power ascension following a mid-cycle maintenance

outage, which required reactor vessel disassembly. Following the outage, the reactor vessel

was reassembled and operators commenced startup operations. With the reactor in Startup

Mode (Mode 2) and at normal operating pressure, operators noted increasing drywell floor drain

leakage. At 3:01 a.m. eastern standard time (EST), an Unusual Event was declared as a result

of unidentified drywell leakage exceeding 10 gallons per minute (gpm). At 3:09 a.m. EST, a

manual reactor scram was initiated from approximately 7 percent of rated thermal power due to

the continued increase in unidentified drywell leakage. Following the scram, the reactor was

depressurized and the unidentified leak rate decreased to less than 10 gpm within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. At

1:45 p.m. EST on November 17, 2011, with the reactor in Cold Shutdown (Mode 4), leak investigation activities determined that the reactor vessel head studs were not fully tensioned

during startup operations; therefore, an unanalyzed condition existed at Brunswick 2.

Subsequently, it was determined that none of the 64 reactor vessel head studs were adequately

tensioned.

Reactor vessel head stud tensioning is accomplished by attaching a tensioning device to the

studs uppermost threads. Hydraulic pressure is applied to the tensioning device, which

stretches the stud. With the stud elongated by the tensioning device, personnel rotate the stud

nut until it makes firm contact with the washer on the head flange. When the hydraulic pressure

is released, the nut maintains the tension and elongation in the stud, applying closure pressure

to the flanges of the reactor vessel and head.

The licensees investigation determined that this event was the result of errors made while

operating the reactor vessel head stud tensioning equipment and during the validation process

to ensure the head was properly tensioned. Following the event, the licensee assessed the

stud tensioning process through equipment troubleshooting, review of the reactor vessel

reassembly procedure (Procedure 0SMP-RPV502), and interviews with refuel floor personnel.

The equipment was found to be fully functional. However, the licensee determined that

personnel operating the stud tensioning equipment misinterpreted the digital display of the

hydraulic pressure being applied to elongate the studs. Specifically, the licensee found that

personnel incorrectly believed that the actual hydraulic pressure being applied to the tensioning

device was a factor of ten greater than the pressure indicated on the device. As a result, none

of the 64 studs were properly tensioned during the reactor vessel assembly process.

The Stud Elongation Measurement System (SEMS III) is used at Brunswick 2 to validate proper

stud elongation. Based on interviews with personnel, the licensee determined that the refuel

floor crew incorrectly concluded that the target stud elongation value of 0.045 inches was

achieved when the elongation values indicated on the SEMS III device were only between

+/-0.004 inches. The licensee attributed this error to the crew incorrectly assuming that the

elongation value of 0.045 inches was automatically deducted from the post-tensioned

elongation indication on the SEMS III device. Furthermore, the elongation values of +/-0.004 inches, as indicated on the SEMS III device, correspond to the stud elongation tolerance

specified in Procedure 0SMP-RPV502. Accordingly, the crew compared the low reading on the

SEMS III device to the stud elongation tolerance in the procedure and erroneously determined

that acceptable stud elongation had been achieved. The quality control inspector concurred

with the consensus opinion of the crew. As a result of these errors, the reactor vessel head

studs were tensioned to only approximately 10 percent of the required amount. Therefore, Brunswick 2 reached Mode 2 with the head not properly tensioned. The increase in leakage

and subsequent reactor scram were a direct result of this condition.

The licensee performed a post-event evaluation of the integrity of the reactor vessel closure

components. The licensee concluded that no reactor coolant pressure boundary components

were damaged or overstressed as result of the event. After completing the integrity evaluation, the reactor vessel was reassembled. Prior to plant restart, a hydrostatic test was completed to

verify that proper head stud tensioning had been achieved. The licensee attributed the root cause of this event to the failure to provide proper training and

lack of procedure guidance to correctly interpret critical data used to validate that the reactor

vessel head studs are properly tensioned. Specifically, the licensee concluded that the operator

errors that occurred during the reactor vessel reassembly evolution were due to an inadequate

understanding of the digital readings displayed on the hydraulic stud tensioning equipment and

the SEMS III stud elongation measurement device. For both cases, the licensee determined

that the crews relied on erroneous assumptions that led to incorrect conclusions.

Licensee Corrective Actions

The licensee revised the reactor vessel reassembly procedure (Procedure 0SMP-RPV502) to

include detailed guidance on the proper use of the SEMS III stud elongation measurement

equipment and the interpretation of hydraulic pressure indications on the stud tensioning device.

The licensee also provided training to refuel floor crew personnel on the proper operation of the

SEMS III and hydraulic stud tensioning equipment during reactor vessel reassembly. The

licensee has revised its refuel floor training and qualification documents to include specific

discussion on the correct operation of the SEMS III equipment and how to properly interpret

hydraulic pressure indications on the stud tensioning device. The licensee has also revised

Procedure 0SMP-RPV502 to require the necessary level of training regarding these activities, as provided in the revised training documents. In addition, the licensee has modified

corporate-wide qualification programs for nuclear fleet refuel floor crew personnel to ensure that

all refuel floor personnel at Brunswick maintain the necessary qualifications for performing their

assigned activities and receive the necessary level of training on the SEMS III and stud

tensioning equipment, as provided in the revised training documents.

The licensee noted that, prior to the event, a decision was made that a post maintenance

reactor vessel pressure test was not necessary because there are no regulatory requirements to

conduct this test following mid-cycle maintenance outages. Therefore, as a corrective action, the licensee revised plant procedures to require a pressure test of the reactor vessel following

mid-cycle maintenance outages that require reactor vessel reassembly.

NRC Special Inspection Team Findings

An NRC special inspection team reviewed the circumstances surrounding this event. The

inspection team reviewed the licensees actions prior to the event and identified examples of

improper procedure adherence that contributed to the inadequate reactor vessel head stud

tensioning. Specifically, the team determined that licensee personnel failed to properly

pressurize the reactor vessel head stud tensioning equipment to the value specified in

Procedure 0SMP-RPV502 because the tensioning equipment operators did not know how to

correctly interpret the hydraulic pressure reading on the tensioning equipment display. The

inspection team also determined that quality control personnel failed to verify proper reactor

vessel stud elongation in accordance with stud elongation values specified in Procedure

0SMP-RPV502. Further, the inspection team determined that nine of the twelve refuel floor

personnel performing reactor vessel reassembly did not have the necessary refuel floor support

training, as required by Procedure TRN-NGCC-1000, Conduct of Training. Finally, based on

its review of Procedure 0PLP-20, Post Maintenance Testing Program, which specifies plant

equipment shall be tested consistent with their safety functions following maintenance activities that may have impaired proper functioning of the components, the inspection team determined

that the licensee failed to specify an adequate post maintenance test to verify the pressure

retaining capability of the reactor vessel following a mid-cycle maintenance outage.

The Brunswick Steam Electric Plant, Unit 2, Licensee Event Report (LER) 50-324/2-2011-002, dated January 16, 2012, contains further discussion of this event. The LER is available on the

NRCs public Web site under Agencywide Documents Access and Management System

(ADAMS) Accession No. ML12031A167. Additional information is available in NRC Special

Inspection Report 05000324/2011013, dated January 25, 2012, under ADAMS Accession

No. ML120250556; and NRC Inspection Report 05000324/2012007, dated April 20, 2012, under

ADAMS Accession No. ML12114A036.

DISCUSSION

Section 50.120, Training and qualification of nuclear power plant personnel, of 10 CFR states, in part, that the training program must incorporate the instructional requirements necessary to

provide qualified personnel to operate and maintain the facility in a safe manner in all modes of

operation. Criterion V, Instructions, Procedures, and Drawings, of Appendix B, Quality

Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants, to 10 CFR Part 50

states, in part, that instructions, procedures, or drawings shall include appropriate quantitative or

qualitative acceptance criteria for determining that important activities have been satisfactorily

accomplished.

The root cause of this event was the failure to provide the necessary training and procedure

guidance to correctly interpret critical indications on the stud tensioning and stud elongation

measurement equipment for verifying that proper stud tensioning had been achieved. The

failure to adequately tension the reactor vessel closure head studs during reactor vessel

reassembly undermined the integrity of the reactor coolant pressure boundary, one of the

primary barriers to fission product release, during startup operations.

In addition, a decision was made that a post maintenance reactor vessel pressure test was not

necessary because there are no regulatory requirements to conduct this test following mid-cycle

maintenance outages. However, the reactor vessel head was removed and reinstalled during

this outage in the same fashion as during a refueling outage. Therefore, this event highlights

the importance of conducting mid-cycle maintenance outage activities, particularly those that

require reactor vessel disassembly and reassembly, with the same level of rigor as scheduled

refueling outage activities.

This event also highlights the importance of human performance and oversight of maintenance

activities. For example, operators of the stud tensioning equipment were not familiar with the

pressure display, yet they proceeded with tensioning based on an incorrect interpretation of

indicated tensioner pressure. In addition, a licensee lead mechanic and a quality control

inspector signed a procedure checklist for stud elongation measurements using flawed data, based on incorrect explanations by other members of the maintenance crew. Other findings

related to human performance can be found in the April 20, 2012, inspection report.

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) or Office of New Reactors project manager.

/RA/

/RA by JLuehman for/

Timothy J. McGinty, Director

Laura A. Dudes, Director

Division of Policy and Rulemaking

Division of Construction Inspection

Office of Nuclear Reactor Regulation

and Operational Programs

Office of New Reactors

Technical Contacts: Christopher R. Sydnor, NRR

Molly J. Keefe, NRR

301-415-6065

301-415-5717 E-mail:

E-mail

Christopher.Sydnor@nrc.gov

Molly.Keefe@nrc.gov

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

ML12264A518 *via e-mail TAC No. ME8863 OFFICE

EVIB:NRR

AHPB:NRR

Tech Editor

BC:EVIB:NRR

BC:AHPB:NRR

NAME

CSydnor*

MKeefe*

CHsu*

SRosenberg*

UShoop*

DATE

11/27/12

11/27/12

10/11/12

11/27/12

11/27/12 OFFICE

D:DE

PM:PGCB:NRR

LA:PGCB:NRR

BC:PGCB:NRR

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NAME

PHiland (MCheok for) TAlexion

CHawes

DPelton (EBowman for) LDudes (JLuehman

for)

DATE

11/27/12*

11/28/12

11/29/12

11/27/12*

12/03/12 OFFICE

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NAME

TMcGinty

OFFICE

12/10 /12