Information Notice 2016-11, Potential for Material Handling Events to Cause Internal Flooding

Potential for Material Handling Events to Cause Internal Flooding

ML16154A022
Person / Time
Issue date:	08/12/2016
From:	Michael Cheok, Louise Lund Office of New Reactors, Office of Nuclear Reactor Regulation
To:
Banic M
References
IN-16-011
Download: ML16154A022 (6)

UNITED STATES

NUCLEAR REGULATORY COMMISSION

OFFICE OF NUCLEAR REACTOR REGULATION

OFFICE OF NEW REACTORS

WASHINGTON, DC 20555-0001 August 12, 2016 NRC INFORMATION NOTICE 2016-11: POTENTIAL FOR MATERIAL HANDLING

EVENTS TO CAUSE INTERNAL FLOODING

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 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 that indicated material handling events could cause

internal flooding that exceeds flood levels considered in the facility design basis. The NRC

expects that recipients will review the information for applicability to their facilities and 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

Fort Calhoun Station

On August 2, 2013, the engineering staff at Fort Calhoun Station (FCS) identified that the

seismic analysis of the intake structure crane did not evaluate the cranes ability to withstand a

seismic event when in use. At the time of discovery, the unit was in cold shutdown. An

investigation identified that the crane had been used when the raw water pumps were required

to be operable. The licensee had previously completed a load drop analysis for the intake

structure and determined that a load drop from the crane would not cause damage to the intake

structure. However, the engineering staff found that potential damage to the unprotected fire

protection headers that exist in the intake structure had not been considered in the load drop

analysis. Because the crane had not been verified to withstand seismic effects, the licensee

concluded that this piping could be damaged by falling equipment if the intake structure crane

was in use during a seismic event. The engineering staff concluded that the volume of flooding

that could be produced by this event was outside of the assumptions of the intake structure

internal flooding analysis and could result in all four raw water pumps becoming inoperable.

ML16154A022 At FCS, the raw water system performs an essential safety function. In combination with the

component cooling water system, the raw water system performs the safety-related

design-basis accident heat removal function and the decay heat removal function. Therefore, the licensee identified this condition as one that could have prevented fulfillment of an essential

safety function.

The safety significance of this identified condition was low. Unlike many U.S. power reactor

facilities, the FCS emergency onsite diesel generators have air-cooled radiators and, therefore, the electric power distribution system does not rely on the raw water system. The licensee also

had an existing abnormal operating procedure specifically developed to address a total loss of

raw water. Furthermore, the primary corrective action for this identified issue was limited to

development of a new seismic analysis that determined the crane could be safely operated with

an attached load during a seismic event.

Additional information is available in Fort Calhoun Station Licensee Event Report (LER)

50-285/2013012, dated September 30, 2013.1 Further information also appears in NRC

Inspection Reports 05000285/2013015 and 05000285/2015007, dated November 8, 2013,2 and

April 16, 2015,3 respectively.

Arkansas Nuclear One, Units 1 and 2

On March 31, 2013, during an Arkansas Nuclear One (ANO) Unit 1 outage, the licensee was

moving the Unit 1 main generator stator out of the turbine building when an inadequately

designed and untested temporary lifting rig collapsed. The collapse caused the 525-ton stator

to fall onto the Unit 1 turbine deck, roll off the damaged deck, and fall approximately 30 feet into

the train bay between Unit 1 and Unit 2. The stator impact caused substantial damage to the

Unit 1 turbine building structure and power distribution systems, and parts of the collapsing lift

rig struck structures and components on the Unit 2 side of the turbine building.

At the time of the event, Unit 1 was shut down in a refueling outage with the reactor vessel head

off and fuel in the vessel. The partial collapse of the turbine deck damaged non-vital electrical

buses supplying offsite power to Unit 1. This damage to the electrical buses resulted in a loss

of normal offsite power to Unit 1 for 6 days, but power was available from emergency diesel

generators to power both trains of safety-related equipment in Unit 1.

Unit 2 was operating at 100 percent power with no major evolutions in progress at the time of

the event. When the temporary lift rig collapsed, components of the lift rig impacted Unit 2 structures and components. The vibration from the impact triggered a relay contact that opened

the breaker supplying power to one of the operating reactor coolant pumps, resulting in an

automatic reactor shutdown. The impact also ruptured an 8-inch fire main. The loss of

pressure in the fire main caused both the normal diesel-driven and a temporary motor-driven fire

pump to start as designed. Operators secured the diesel-driven pump within 15 minutes, but

temporary motor-driven fire pump operation continued for over 40 minutes, as indicated by flow

from the rupture. Although much of the water flowed out of the turbine building train bay, water

from the rupture also flowed to areas of the turbine and auxiliary buildings, causing additional

damage. The damage included loss of one offsite source to Unit 2 after water caused an

1 Agencywide Documents Access and Management System (ADAMS) Accession No. ML13274A638. ADAMS is

accessible through the NRCs public Web site at http://www.nrc.gov, under NRC Library.

2 ADAMS Accession No. ML13312A876.

3 ADAMS Accession No. ML15106A891. electrical fault inside the Unit 2 non-safety switchgear in the turbine building approximately

33 minutes after the collapse.

The NRC staff determined that the temporary lift rig collapse had substantial safety significance

for both Unit 1 and Unit 2. The staff found that the direct and indirect damage caused to the

electrical distribution system and the complications associated with water around the switchgear

would have posed significant challenges to recovery of offsite power, if the onsite sources had

not functioned properly. Corrective actions included: (1) modifying procedures related to

handling of heavy loads; (2) training the facility staff on the revised requirements for handling

heavy loads; and (3) repairing the damaged Unit 1 turbine structure, fire main system, and both

Unit 1 and Unit 2 electrical systems.

The NRC dispatched an augmented inspection team to review the facts surrounding the event, as documented in Arkansas Nuclear OneNRC Augmented Inspection Team Report

05000313/2013011 and 05000368/2013011, dated June 7, 2013.4 Additional information is

available in Arkansas Nuclear OneNRC Augmented Inspection Team Follow-Up Inspection

Report 05000313/2013012 and 05000368/2013012; Preliminary Red and Yellow Findings, dated March 24, 2014.5 Further information is available in Arkansas Nuclear One, Units 1 and 2, LER 50-313/2013-001-00, dated May 24, 2013,6 and Supplemental LER

50-313/2013-001-01, dated August 22, 2013.7

BACKGROUND

Related NRC Regulations

The regulations in 10 CFR 50.65, Requirements for monitoring the effectiveness of

maintenance at nuclear power plants, paragraph (a)(4), require that each licensee assess and

manage the increase in risk that may result from proposed maintenance activities. The scope of

this assessment may be limited to structures, systems, and components (SSCs) that a

risk-informed evaluation process has shown to be significant to public health and safety.

DISCUSSION

Licensees commonly undertake activities involving movement of heavy components, particularly

activities supporting refueling and refurbishment of large plant components. One class of these

activities involves refueling and refurbishment of large components using permanently installed

cranes that were evaluated as part of each licensees heavy load handling program. A second

class of activities consists of less frequent heavy load movements for maintenance and

refurbishment that had not been considered in the scope of the heavy load handling program.

This second class of load movements may involve the use of permanently installed cranes that

have not been evaluated for use under all plant conditions or temporary overhead handling

equipment. The events identified at FCS and ANO are examples of this second class of

activities. These heavy load movements may be subject to the requirements of

10 CFR 50.65(a)(4) to assess and manage the risk of heavy load movements associated with

maintenance activities.

4 ADAMS Accession No. ML13158A242.

5 ADAMS Accession No. ML14083A409.

6 ADAMS Accession No. ML13144A220.

7 ADAMS Accession No. ML13234A241. Material handling activities evaluated within the scope of each licensees heavy load handling

program have included consideration of the overhead handling system design, testing, and

maintenance. The evaluation of the design, testing, and maintenance applied to these handling

systems provides assurance that the structure of the handling system is robust. Consequently, traditional load drop analyses only postulated failures in the hoisting machinery and rigging, which would limit potential effects to equipment near the load. However, use of temporary

overhead handling systems or permanent overhead handling systems not previously evaluated

for use under all plant conditions may not provide the same level assurance in the structural

integrity of the system. Therefore, licensees may wish to consider the potential for structural

failures and consequential plant damage when assessing measures to manage the risk of

heavy load movements using these types of overhead handling systems for maintenance

related activities.

Handling system structural failures could affect SSCs well away from the load itself because the

overhead handling system structure often spans long distances. Although load drop analyses

typically evaluate the effect on SSCs immediately below the load path, the events at FCS and

ANO demonstrate the potential for damage to SSCs separate from the load path. The FCS

event report addressed potential damage to piping systems under the overhead crane bridge

that the licensee did not consider in the completed load drop analysis, because the piping was

not under the load. Similarly, the temporary handling system collapse at ANO damaged fire

protection system piping outside the footprint of the handling system structure. The

consequences of a material handling accident can be magnified by potential internal flooding

because flooding from pipe breaks can:

• be of greater magnitude than that considered in the design basis,

• propagate to other areas, and

• affect redundant components.

For these cases, licensees may wish to manage the increase in risk associated with the

maintenance activity by enhancing the qualification of the handling system structure, as

completed at FCS, or evaluating the effects of structural failures on equipment beyond the

immediate vicinity of the load.

CONTACT

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

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

Regulation or Office of New Reactors project manager.

/ra/ /ra/ (Mirela Gavrilas for)

Michael C. Cheok, Director Louise Lund, Director

Division of Construction Inspection Division of Policy and Rulemaking

and Operational Programs Office of Nuclear Reactor Regulation

Steve Jones, NRR

301-415-2712 E-mail: Steve.Jones@nrc.gov

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

CONTACT

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

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

Regulation or Office of New Reactors project manager.

/ra/ /ra/ (Mirela Gavrilas for)

Michael C. Cheok, Director Louise Lund, Director

Division of Construction Inspection Division of Policy and Rulemaking

and Operational Programs Office of Nuclear Reactor Regulation

Steve Jones, NRR

301-415-2712 E-mail: Steve.Jones@nrc.gov

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

ADAMS Accession Number: ML16154A022 *via e-mail TAC MF7539 OFFICE NRR/DSS/SBPB/TL* TECH EDITOR* NRR/DPR/PGCB/LA* NRR/DSS/SBPB/BC* NRR/DSS/D*

NAME SJones JDougherty ABaxter RDennig TMcGinty

DATE 05/23/16 05/20/16 07/20/16 07/21/16 07/21/16 OFFICE NRR/DPR/PGCB/PM NRO/DSRA/BC* NRO/DCIP/BC* NRR/DPR/PGCB/LA* NRR/DPR/PGCB/BC

NAME MBanic ADias RLukes ELee SStuchell

DATE 08/02/16 07/29/16 07/22/16 08/02/16 08/02/16 OFFICE NRO/DCIP/D NRR/DPR/D

NAME MCheok LLund

(MGavrilas for

w/comments)

DATE 08/08/16 08/12/16 OFFICIAL RECORD COPY