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NRC INFORMATION NOTICE 2016-11: POTENTIAL FOR MATERIAL HANDLING EVENTS TO CAUSE INTERNAL FLOODING | 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== | ==ADDRESSEES== | ||
All holders of an operating license or construction permit for a nuclear power reactor under | All holders of an operating license or construction permit for a nuclear power reactor under | ||
Code of Federal Regulations (10 CFR) Part 50, | 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 | |||
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, | 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 | 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== | ==DESCRIPTION OF CIRCUMSTANCES== | ||
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 | ===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. | 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 | investigation identified that the crane had been used when the raw water pumps were required | ||
to be operable. | 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 | protection headers that exist in the intake structure had not been considered in the load drop | ||
analysis. | 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 | concluded that this piping could be damaged by falling equipment if the intake structure crane | ||
was in use during a seismic event. | 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. | safety function. | ||
The safety significance of this identified condition was low. | 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 | 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. | 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 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 | Unit 1 turbine building structure and power distribution systems, and parts of the collapsing lift | ||
Line 93: | Line 125: | ||
At the time of the event, Unit 1 was shut down in a refueling outage with the reactor vessel head | 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. | off and fuel in the vessel. The partial collapse of the turbine deck damaged non-vital electrical | ||
buses supplying offsite power to Unit 1. | 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 | of normal offsite power to Unit 1 for 6 days, but power was available from emergency diesel | ||
Line 103: | Line 135: | ||
Unit 2 was operating at 100 percent power with no major evolutions in progress at the time of | Unit 2 was operating at 100 percent power with no major evolutions in progress at the time of | ||
the event. | 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 | |||
pump to start as designed. | 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 | temporary motor-driven fire pump operation continued for over 40 minutes, as indicated by flow | ||
from the rupture. | 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. | 2 ADAMS Accession No. ML13312A876. | ||
3 ADAMS Accession No. ML15106A891. | 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 | The NRC staff determined that the temporary lift rig collapse had substantial safety significance | ||
for both Unit 1 and Unit 2. | 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 | 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. | 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 | heavy loads; and (3) repairing the damaged Unit 1 turbine structure, fire main system, and both | ||
Line 135: | Line 179: | ||
Unit 1 and Unit 2 electrical systems. | Unit 1 and Unit 2 electrical systems. | ||
The NRC dispatched an augmented inspection team to review | 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, | 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 | ||
dated March 24, 2014. | |||
50-313/2013-001-01, dated August 22, 2013.7 | |||
==BACKGROUND== | |||
== | ===Related NRC Regulations=== | ||
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== | ==DISCUSSION== | ||
Licensees commonly undertake activities involving movement of heavy components, particularly activities supporting refueling and refurbishment of large plant components. | 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 | 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. | 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. | 4 ADAMS Accession No. ML13158A242. | ||
5 ADAMS Accession No. ML14083A409. | 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 | 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. | 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. | 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 | consequences of a material handling accident can be magnified by potential internal flooding | ||
because flooding from pipe breaks can: | 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. | immediate vicinity of the load. | ||
IN | ==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 | |||
== | ===Office of New Reactors=== | ||
===Technical Contact:=== | |||
===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 | |||
Office of | 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 | |||
===Office of New Reactors=== | |||
== | ===Technical Contact:=== | ||
===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}} | |||
{{Information notice-Nav}} | {{Information notice-Nav}} |
Latest revision as of 19:22, 30 October 2019
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
Office of New Reactors
Technical Contact:
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
Office of New Reactors
Technical Contact:
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