Information Notice 2016-11, Potential for Material Handling Events to Cause Internal Flooding
| ML16154A022 | |
| Person / Time | |
|---|---|
| Issue date: | 08/12/2016 |
| From: | Cheok M C, Louise Lund Office of New Reactors, Office of Nuclear Reactor Regulation |
| To: | |
| Banic M | |
| References | |
| IN-16-011 | |
| Download: ML16154A022 (6) | |
ML16154A022 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 vesse 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 basi The NRC expects that recipients will review the information for applicability to their facilities and consider actions, as appropriate, to avoid similar problem However, suggestions contained in this IN are not NRC requirements; therefore, no specific action or written response is require
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 crane's ability to withstand a seismic event when in us At the time of discovery, the unit was in cold shutdow An investigation identified that the crane had been used when the raw water pumps were required to be operabl 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 structur 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 analysi 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 even 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 inoperabl At FCS, the raw water system performs an essential safety functio 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 functio Therefore, the licensee identified this condition as one that could have prevented fulfillment of an essential safety functio The safety significance of this identified condition was lo 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 syste The licensee also had an existing abnormal operating procedure specifically developed to address a total loss of raw wate 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 even Additional information is available in Fort Calhoun Station Licensee Event Report (LER) 50-285/2013012, dated September 30, 201 Further information also appears in NRC Inspection Reports 05000285/2013015 and 05000285/2015007, dated November 8, 2013,2 and April 16, 2015,3 respectivel 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 collapse 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 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 buildin 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 vesse The partial collapse of the turbine deck damaged non-vital electrical buses supplying offsite power to Unit 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 Unit 2 was operating at 100 percent power with no major evolutions in progress at the time of the even When the temporary lift rig collapsed, components of the lift rig impacted Unit 2 structures and component 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 shutdow The impact also ruptured an 8-inch fire mai The loss of pressure in the fire main caused both the normal diesel-driven and a temporary motor-driven fire pump to start as designe 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 ruptur 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 damag 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. ML13274A63 ADAMS is accessible through the NRC's public Web site at http://www.nrc.gov, under NRC Library. 2 ADAMS Accession No. ML13312A876. 3 ADAMS Accession No. ML15106A89 electrical fault inside the Unit 2 non-safety switchgear in the turbine building approximately 33 minutes after the collaps The NRC staff determined that the temporary lift rig collapse had substantial safety significance for both Unit 1 and Unit 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 properl 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 system The NRC dispatched an augmented inspection team to review the facts surrounding the event, as documented in "Arkansas Nuclear One-NRC Augmented Inspection Team Report 05000313/2013011 and 05000368/2013011," dated June 7, 201 Additional information is available in "Arkansas Nuclear One-NRC Augmented Inspection Team Follow-Up Inspection Report 05000313/2013012 and 05000368/2013012; Preliminary Red and Yellow Findings," dated March 24, 201 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, 201
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 activitie 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 safet
DISCUSSION
Licensees commonly undertake activities involving movement of heavy components, particularly activities supporting refueling and refurbishment of large plant component One class of these activities involves refueling and refurbishment of large components using permanently installed cranes that were evaluated as part of each licensee's heavy load handling progra 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 progra 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 equipmen The events identified at FCS and ANO are examples of this second class of activitie 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 activitie ADAMS Accession No. ML13158A242. 5 ADAMS Accession No. ML14083A40 ADAMS Accession No. ML13144A220. 7 ADAMS Accession No. ML13234A24 Material handling activities evaluated within the scope of each licensee's heavy load handling program have included consideration of the overhead handling system design, testing, and maintenanc The evaluation of the design, testing, and maintenance applied to these handling systems provides assurance that the structure of the handling system is robus Consequently, traditional load drop analyses only postulated failures in the hoisting machinery and rigging, which would limit potential effects to equipment near the loa 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 syste 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 activitie Handling system structural failures could affect SSCs well away from the load itself because the overhead handling system structure often spans long distance 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 pat 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 loa Similarly, the temporary handling system collapse at ANO damaged fire protection system piping outside the footprint of the handling system structur 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 component 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 loa
CONTACT
This IN requires no specific action or written respons 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 manage /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 Librar
CONTACT
This IN requires no specific action or written respons 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 manage /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 Librar 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