ML020220333
| ML020220333 | |
| Person / Time | |
|---|---|
| Site: | Dresden, Browns Ferry, Oconee, Palo Verde, Oyster Creek, Grand Gulf, Limerick, Diablo Canyon, Comanche Peak |
| Issue date: | 01/17/2002 |
| From: | Eltawila F NRC/RES/DSARE |
| To: | Zwolinski J NRC/NRR/DLPM |
| References | |
| Download: ML020220333 (104) | |
Text
January 17, 2002 MEMORANDUM TO:
John A. Zwolinski, Director Division of Licensing Project Management Office of Nuclear Reactor Regulation FROM:
Farouk Eltawila, Director
/RA/
Division of Systems Analysis and Regulatory Effectiveness Office of Nuclear Regulatory Research
SUBJECT:
DRAFT NUREG-XXXX, TECHNICAL ASSESSMENT, GENERIC ISSUE 186: POTENTIAL RISK AND CONSEQUENCES OF HEAVY LOAD DROPS IN NUCLEAR POWER PLANTS Per discussion with Mr. Ray Wharton during a meeting held on December 12, 2001, please distribute the attached draft NUREG to the following nine nuclear facilities using standard distribution lists: (1) Browns Ferry, (2) Comanche Peak, (3) Diablo Canyon, (4) Dresden, (5) Grand Gulf, (6) Limerick, (7) Oconee, (8) Oyster Creek, and (9) Palo Verde. Each of the nine facilities provided crane design and operating data which is referenced in the draft NUREG. The draft NUREG has been revised from an earlier version that you received in a memorandum dated September 19, 2001. Each facility needs to check for the accuracy of the data and information it provided. Beyond this, there are no other requests of the facilities.
This study was initiated in response to candidate generic issue 186, Potential Risk and Consequences of Heavy Load Drops in Nuclear Power Plants. The attached report documents the results of a systematic and comprehensive study of crane operating experience at U.S.
nuclear power plants.
Following a standard RES peer review of the draft NUREG, the Office of Nuclear Regulatory Research (RES) will make appropriate changes to the document; submit it for publication, and provide recommendations to the NRC regarding any necessary actions to be taken in addressing Generic Issue 186. It is our understanding that NRR will coordinate the review by and receipt of comments from the regions.
Comments on the draft NUREG should be provided to Ronald Lloyd of my staff by March 15, 2002.
Attachment:
As stated cc: See attached list
January 17, 2002 MEMORANDUM TO:
John A. Zwolinski, Director Division of Licensing Project Management Office of Nuclear Reactor Regulation FROM:
Farouk Eltawila, Director
/RA/
Division of Systems Analysis and Regulatory Effectiveness Office of Nuclear Regulatory Research
SUBJECT:
DRAFT NUREG-XXXX, TECHNICAL ASSESSMENT, GENERIC ISSUE 186: POTENTIAL RISK AND CONSEQUENCES OF HEAVY LOAD DROPS IN NUCLEAR POWER PLANTS Per discussion with Mr. Ray Wharton during a meeting held on December 12, 2001, please distribute the attached draft NUREG to the following nine nuclear facilities using standard distribution lists: (1) Browns Ferry, (2) Comanche Peak, (3) Diablo Canyon, (4) Dresden, (5) Grand Gulf, (6) Limerick, (7) Oconee, (8) Oyster Creek, and (9) Palo Verde. Each of the nine facilities provided crane design and operating data which is referenced in the draft NUREG. The draft NUREG has been revised from an earlier version that you received in a memorandum dated September 19, 2001. Each facility needs to check for the accuracy of the data and information it provided. Beyond this, there are no other requests of the facilities.
This study was initiated in response to candidate generic issue 186, Potential Risk and Consequences of Heavy Load Drops in Nuclear Power Plants. The attached report documents the results of a systematic and comprehensive study of crane operating experience at U.S.
nuclear power plants.
Following a standard RES peer review of the draft NUREG, the Office of Nuclear Regulatory Research (RES) will make appropriate changes to the document; submit it for publication, and provide recommendations to the NRC regarding any necessary actions to be taken in addressing Generic Issue 186. It is our understanding that NRR will coordinate the review by and receipt of comments from the regions.
Comments on the draft NUREG should be provided to Ronald Lloyd of my staff by March 15, 2002.
Attachment:
As stated cc: See attached list Distribution w/att.:
REAHFB R/F CPaperiello, DEDMRS DSARE R/F WKane, DEDR AThadani/RZimmerman, RES JShea, OEDO PNorian, RES ISchoenfeld, OEDO CAder, RES RSavio, ACRS C:\\Program Files\\Adobe\\Acrobat 4.0\\PDF Output\\HeavyLoa.wpd
- See previous concurrence OAR in ADAMS? (Y or N) Y ADAMS ACCESSION NO.: ML020220333 TEMPLATE NO. RES-006 Publicly Available? (Y or N) N DATE OF RELEASE TO PUBLIC SENSITIVE? N To receive a copy of this document, indicate in the box: "C" = Copy without enclosures "E" = Copy with enclosures "N" = No copy OFFICE REAHFB REAHFB C:REAHFB D:DSARE NAME RLloyd:mtb HVanderMolen JFlack by JIbarra FEltawila DATE 01/15/02*
01/15/02*
01/14/02*
01/17/02*
January 17, 2002 J. Zwolinski, et.al.
2 Memorandum Dated:
SUBJECT:
DRAFT NUREG-XXXX, TECHNICAL ASSESSMENT, GENERIC ISSUE 186: POTENTIAL RISK AND CONSEQUENCES OF HEAVY LOAD DROPS IN NUCLEAR POWER PLANTS cc w/atts.:
M. Mayfield, RES S. Newberry, RES W. Dean, NRR L. Plisco, R-II G. Grant, R-III K. Brockman, R-IV W. Lanning, R-I B. Mallett, R-II G. Grobe, R-III A. Howell, R-IV cc w/o atts.:
B. Sheron, NRR D. Matthews, NRR G. Holahan, NRR J. Strosnider, NRR B. Boger, NRR
NUREG-XXXX Technical Assessment Generic Issue 186:
Potential Risk and Consequences of Heavy Load Drops in Nuclear Power Plants Prepared by:
R.L. Lloyd Division of Systems Analysis and Regulatory Effectiveness Office of Nuclear Regulatory Research U.S. Nuclear Regulatory Commission Washington, DC 20555-001
iii ABSTRACT This report was written in response to a candidate generic issue 186, Potential Risk and Consequences of Heavy Load Drops in Nuclear Power Plants, to determine the likelihood and significance of very heavy load drops (i.e., loads of approximately 30 tons or greater). This report describes the results of a detailed review of crane operating experience at U.S. nuclear power plants from 1968 through 1999. Crane operating experience information came from several sources including; actual crane operating experience from U.S. nuclear power plants, licensee event reports (10 CFR 50.72 and 10 CFR 50.73), NRC inspection reports, licensee correspondence, and crane vendor reports. This report lists the causes and results of documented crane issues, and estimates the probabilities of selected load drop events. In addition, major crane operating experience reports issued by the New Mexico Environmental Evaluation Group, the Department of Energy, the Department of the Navy, the California Division of Occupational Safety and Health, and Appendix A to NUREG-1738 titled Structural Integrity of Spent Fuel Pool Structures Subject to Heavy Loads Drops, have been reviewed to provide additional insights.
v CONTENTS ABSTRACT............................................................... iii EXECUTIVE
SUMMARY
..................................................... ix ACKNOWLEDGMENTS..................................................... xi ABBREVIATIONS.........................................................
xiii 1
INTRODUCTION...................................................... 1 1.1 Background.................................................... 1 1.2 Definitions..................................................... 1 1.3 Precursors to Initiation of Generic Issue 186........................... 3 2
CRANE OPERATING EXPERIENCE AT U.S. NUCLEAR POWER PLANTS........ 5 2.1 Crane Event Database Categories and Subcategories................... 6 2.2 Analysis of Documented Crane Issues
............................... 6 2.2.1 Reported Crane Issues...................................... 7 2.2.2 Crane Reports Due to Not Following Procedures.................. 8 2.2.3 Crane Issue Distribution by Crane Type......................... 9 2.2.4 Crane Issues Due to Hardware Deficiencies.................... 10 2.2.5 Crane Issues Caused by Programmatic Issues.................. 10 2.2.6 Safety Implication of Crane Issues............................ 12 2.2.7 Crane Type Involved in Load Slip or Drop...................... 13 2.2.8 Distribution of Load Slips and Drops (1968-1999)................ 18 2.2.9 Crane Events Resulting in Deaths or Injuries.................... 19 2.2.10 Description and Distribution of Crane Related Deaths and Injuries
... 20 2.2.11 Distribution of Crane Issues by Facility......................... 23 3
LICENSEE CRANE OPERATING EXPERIENCE AT OPERATING NUCLEAR POWER PLANT FACILITIES (POST NUREG-0612)................................. 24 3.1 Pilot Plants for Crane Program and Operating Experience Reviews........ 24 3.2 Crane Operating Experience at Pilot Plants........................... 25 3.3 Estimated Crane Operating Experience at US Nuclear Power Plants (Post NUREG-0612)................................................. 26 3.4 Very Heavy Load Slips and Drops at Operating Facilities (Post NUREG-0612)
............................................................ 26 3.5 Crane Load Drop Event Tree...................................... 27 3.6 No Accident Sequence Precursor Events Involving Cranes............... 30 4
LICENSEE LOAD DROP CALCULATION METHODOLOGY AND RESULTS
...... 32 5
GUIDANCE FOR HEAVY LOAD MOVEMENTS AND CRANE CLASSIFICATION... 48 5.1 Guidance from NUREG-0612
..................................... 48 5.2 Licensee Response to NRC Bulletin 96-02........................... 49 5.3 Single-failure-proof Crane Classification............................. 50
vi CONTENTS (continued) 6 CRANE OPERATING EXPERIENCE STUDIES............................. 53 6.1 NUREG-0612, Control of Heavy Loads at Nuclear Power Plants........... 53 6.2 EEG-74, Probability of Failure of the TRUDOCK Crane System at the Waste Isolation Pilot Plant (WIPP)....................................... 54 6.3 Department of Energy Study, Independent Oversight Special Study of Hoisting and Rigging Incidents Within the Department of Energy................. 54 6.4 California Department of Industrial Relations, Crane Accidents 1997 - 1999.. 55 6.5 NUREG-1738, Technical Study of Spent Fuel Pool Accident Risk at Decommissioning Nuclear Power Plants............................. 55 6.6 Navy Crane Events............................................. 56 7
REFERENCES...................................................... 57 APPENDICES Appendix A Crane Issues at U.S. Nuclear Power Plants 1968 through October 1999 Appendix B EEG-74: Probability of Failure of the TRUDOCK Crane System at the Waste Isolation Pilot Plant (WIPP), May 2000 Appendix C Independent Oversight Special Study of Hoisting and Rigging Incidents Within the Department of Energy, October 1996 Appendix D Crane Accidents 1997 - 1999, May 2000 Appendix E NUREG-1738, Technical Study of Spent Fuel Pool Accident Risk at Decommissioning Nuclear Power Plants, Appendix A2C, February 2001 Appendix F Navy Crane Operating Experience TABLES Table 1: Crane event database categories and subcategories Table 2: Reported crane issues involving a load drop or load slip Table 3: Deaths associated with crane operation Table 4: Pilot plants for crane program and operational experience reviews Table 5: Total number of lifts with loads of approximately 30 tons or greater Table 6: Load slips and drops of very heavy loads occurring at operating nuclear facilities (Post NUREG-0612)
Table 7: Crane events resulting in a loss or partial loss of offsite power Table 8: Load drop calculations for very heavy components Table 9: Dresden reactor building steel superstructure interaction summary Table 10: Root causes of crane incidents at DOE facilities
vii FIGURES Figure 1: Documented crane issues at U.S. nuclear power plants (1968-1999)
Figure 2: Crane issues caused by not following procedures (1968-1999)
Figure 3: Crane issue distribution by crane type Figure 4: Crane issues due to hardware deficiencies Figure 5: Program Implementation issues (1968-1999)
Figure 6: Number of crane issues by category (1968-1999)
Figure 7: Crane type involved in load slip or drop (1968-1999)
Figure 8: Distribution of load slips and drops (1968-1999)
Figure 9: Crane types involved in deaths or injuries (1968-1999)
Figure 10: Distribution of crane related deaths and injuries (1968-1999)
Figure 11: Distribution of crane issues by facility, on a per unit basis (1968-1999)
Figure 12: Load drop event tree
ix EXECUTIVE
SUMMARY
This NUREG forms the technical basis for any recommendations or corrective actions regarding the Technical Assessment of Candidate Generic Issue 186, Potential Risk and Consequences of Heavy Load Drops in Nuclear Power Plants (To Be Written)
xi ACKNOWLEDGMENTS This report benefitted greatly through the willingness of several staff at Browns Ferry, Comanche Peak, Diablo Canyon, Dresden, Grand Gulf, Limerick, Oconee, Oyster Creek, and Palo Verde nuclear power plants, in researching, documenting, and sharing crane operating experience information. Appreciation is also expressed to those individuals in industry, the public, and the U.S. Nuclear Regulatory Commission who reviewed this report for accuracy.
xiii ABBREVIATIONS ACI American Concrete Institute AE architect engineer AIT Augmented Inspection Team ASP accident sequence precursor BWR boiling water reactor CCDP conditional core damage probability CFR Code of Federal Regulations CRDM control rod drive mechanism DOE U.S. Department of Energy EDG emergency diesel generator EEG Environmental Evaluation Group EQE EQE International FSAR final safety analysis report GI generic issue GL generic letter HT height HVAC heating, ventilation, and air conditioning IC Interaction Coefficient IMIS integrated management information system IN information notice ISFSI independent spent fuel storage installation NEI Nuclear Energy Institute NOG nuclear overhead and gantry NRC U.S. Nuclear Regulatory Commission NRR Office of Nuclear Reactor Regulation (NRC)
NSSS nuclear steam supply system NUDOCS nuclear documents system OBE operating basis earthquake OMDS Office of Management Data Services PWR pressurized water reactor RC reinforced concrete RCP reactor coolant pump RES Office of Nuclear Regulatory Research (NRC)
xiv ABBREVIATIONS (Continued)
RPV reactor pressure vessel RWCU reactor water clean up SFP spent fuel pool SSE safe shutdown equipment SWEC Stone and Webster Engineering Company UBC Uniform Building Code UFSAR Updated Final Safety Analysis Report USI unresolved safety issue WIPP Waste Isolation Pilot Plant WT weight
1 1
INTRODUCTION
1.1 Background
In nuclear plant operation, maintenance and refueling activities, heavy loads may be handled in several plant areas. If these loads were to drop they could impact on stored spent fuel, fuel in the core, or on equipment that may be required to achieve safe shutdown or permit continued decay heat removal. In some instances, load drops at specific times and locations, could potentially lead to offsite doses that exceed 10 CFR Part 100 limits.
In April 1999, a candidate generic issue (GI) was proposed (Ref. 1) by the Office of Nuclear Reactor Regulation (NRR) of the U.S. Nuclear Regulatory Commission (NRC). NRR requested the Office of Nuclear Regulatory Research (RES) within the NRC to evaluate the issue. NRR was concerned that although licensees may be operating within the regulatory guidelines in Generic Letter (GL) 85-11, Completion of Phase II of Control of Heavy Loads at Nuclear Power Plants NUREG-0612, they may not be taking action above and beyond existing regulations to maintain the level of protection necessary to avoid undue risk to public health and safety.
In May 1999, RES informed NRR (Ref. 2) that the candidate GI was accepted, and was given the title GI-186, Potential Risk and Consequences of Heavy Load Drops in Nuclear Power Plants. Ref. 2 indicated that GI-186 would be prioritized in accordance with RES Office Letter No. 7, Procedure for Identification, Prioritization, Resolution, and Tracking of Generic Issues.
With the advent of NRC Management Directive 6.4, Generic Issue Program, in July 1999, it was decided to process this new issue in accordance with MD 6.4 instead of Office Letter No. 7.
1.2 Definitions Auxiliary hoist. Supplemental hoisting unit usually of lower load rating and higher speed than the main hoist.
Bridge. That part of a crane consisting of one or more girder, trucks, end ties, footwalks, and drive mechanism, which carries the trolley or trolleys.
Crane. A machine for lifting and lowering a load and moving it horizontally, with the hoisting mechanism and integral part of the machine.
Critical load. Any load that, if dropped, could be the direct or indirect cause of release of radioactivity.
Drum. The cylindrical member around which the ropes are wound for lifting or lowering the load.
Gantry crane. A crane similar to an overhead crane except that the bridge for carrying the trolley or trolleys is rigidly supported on two or more legs running on fixed rails or other runway.
Handling system. All load bearing components used to lift the load, including the crane or hoist, the lifting device, and interfacing load lift points.
2 Heavy load. Any load, carried in a given area after a plant becomes operational that weighs more than the combined weight of a single spent fuel assembly and its associated handling tool for the specific plant in question.
Hoist. A machinery unit that is use for lifting or lowering a freely suspended (unguided) load.
Lifting devices. Devices that are not reeved onto the hoist ropes, such as hook-on buckets, magnets, grabs and other supplemental devices used for ease of handling certain types of loads. The weight of these devices is to be considered part of the rated load.
Load. The total superimposed weight on the load block or hook.
Load block. The assembly of hook or shackle, swivel, bearing, sheaves, pins, and frame suspended by the hoisting rope or load chain.
Load drop. A situation where the load may descend uncontrollably, but impacts other equipment and does damage.
Load slip. A situation where the load may descend uncontrollably, but come to a stop without impacting or damaging other equipment.
Load hang-up. The act in which the load block and/or load is stopped by a fixed object during hoisting, thereby possibly overloading the hoisting system.
Main hoist. The primary hoist mechanism provided for lifting and lowering the rated load.
Overhead crane. A crane with a single-or multiple-girder movable bridge carrying a moveable or fixed hoisting mechanism and traveling on an overhead fixed runway structure.
Overload. Any load greater than the rated load.
Pendant station. Controls suspended from the crane for operating the unit from the floor.
Polar crane. An overhead or gantry crane that travels on a circular runway.
Reeving. A system in which a rope travels around drums, or sheaves.
Rope. Refers to wire rope unless otherwise specified.
Safe load travel path. A path defined for transport of a heavy load that will minimize adverse effects, if the load is dropped, in terms of releases of radioactive material and damage to safety systems. This path should be administratively controlled by procedure and/or clearly outlined by markings on the floor where the load is to be handled. It may also be enforced by mechanical stops and/or electrical interlocks.
Safe shutdown equipment. Safety related equipment and associated subsystems that would be required to bring the plant to cold shutdown conditions or provide continued decay heat removal following the dropping of heavy load. Safety functions that should be preserved are: to maintain
3 reactor coolant pressure boundary; capability to reach and maintain subcriticality; removal of decay heat; and to maintain integrity of components whose failure could result in excessive offsite releases.
Sheave. A grooved wheel or pulley used with a rope to change direction and point of application of a pulling force.
Single-failure-proof crane. When reliance for the safe handling of critical loads is placed on the crane system itself, the system should be designed so that a single failure will not result in the loss of the capability of the system to safely retain the load. These features are limited to the hoisting system and to braking systems for trolley and bridge. Other load-bearing items such as girders should be conservatively designed but need not be considered single failure proof.
(See NUREG-0554, Single-Failure-Proof Cranes for Nuclear Power Plants.)
Special lifting devices. A lifting device that is designed specifically for handling a certain load or loads, such as the lifting rigs for the reactor vessel head or vessel internals, or the lifting device for a spent fuel cask.
Spent fuel. Fuel that has been critical in the core and is considered no longer sufficiently active to be of use in powering the reactor and therefore is soon to be, or already has been, removed from the reactor.
Trolley. The unit that travels on the bridge rails and supports the load block.
Truck. A unit consisting of a frame, wheels, bearings, and axles that supports the bridge girders, the end ties of an overhead crane, or the sill of a gantry crane.
Two-blocking. The act of continued hoisting to the extent that the upper head block and the load block are brought into contact, and, unless additional measures are taken to prevent further movement of the load block, excessive loads will be created in the rope reeving system, wit the potential for rope failure and dropping of the load.
Very heavy load. Any load weighing approximately 30 tons or more.
1.3 Precursors to Initiation of Generic Issue 186 Several related events took place that led up to the initiation of GI-186. Significant related documents are discussed in chronological order.
Unresolved Safety Issue (USI) A-36, Control of Heavy Loads near Spent Fuel (1970s)
This issue focused mainly on potential consequences of a heavy load drop on fuel assemblies in either the spent fuel pool area or in the reactor, that may result in; (1) a release of radioactivity because of a cladding breach, or (2) a critical mass of fuel in the core or in the spent fuel pool. USI A-36 was resolved with the issuance of NUREG-0612, Control of Heavy Loads at Nuclear Power Plants, and revisions to Section 9.1.5 of the Standard Review Plan, Overhead Heavy Load Handling Systems.
4 NUREG-0554, Single-Failure-Proof Cranes for Nuclear Power Plants (May 1979)
NUREG-0554 was developed to provide design, installation, testing and quality assurance requirements for single-failure-proof cranes. The NRC has licensed reactors on the basis that the safe handling of critical loads can be accomplished by adding safety features to the handling equipment, by adding special features to the structures and areas over which the critical load is carried, or by a combination of the two. When reliance for the safe handling of critical loads is placed on the crane system itself, the system should be designed so that a single failure will not result in the loss of the capability of the system to safely retain the load. This document (Ref. 3) identifies features of the design, fabrication, installation, inspection, testing, and operations of single-failure-proof overhead crane handling systems (limited to the hoisting system and to braking systems for trolley and bridge).
NUREG-0612, Control of Heavy Loads at Nuclear Power Plants (July 1980)
This report (Ref. 4) provides the results of the review of the handling of heavy loads and includes the task groups recommendations on actions that should be taken to assure safe handling of heavy loads. This report completed Task A-36 described earlier.
Subsequent documentation divided the NUREG action items into what became known as Phase I (Section 5.1.1) and Phase II (Sections 5.1.2 through 5.1.6). Phase I addresses safe load paths, procedures, crane operator training, special lifting devices, lifting devises that are not specially designed, and crane inspection and maintenance, while Phase II addresses alternative design requirements for cranes located in the spent fuel pool area for Pressurized water reactors (PWRs), the containment building for PWRs, the reactor building for boiling water reactors (BWRs), and in other plant areas for either a PWR or BWR.
Generic Letter 80-113 (originally unnumbered), Control of Heavy Loads, (December 1980)
Generic Letter (GL)80-113 requested that licensees review their controls for handling of heavy loads to determine the extent to which the guidelines of NUREG-0612 are present at their facilities, and to identify the changes and modifications that would be required in order to fully satisfy these guidelines.
Generic Letter 81-07, Control of Heavy Loads (February 3, 1981)
GL-81-07 clarifies parts of GL-80-113 and requests that additional information be provided for analyses. Licensees were requested to provide additional items such as initial conditions/assumptions of postulated load drops, methods used in the analysis, an analysis that demonstrates that ceilings are not penetrated, and an analysis to demonstrate that post-accident dose will be well within 10 CFR Part 100 limits.
5 Generic Letter 85-11, Completion of Phase II of Control of Heavy Loads at Nuclear Power Plants, NUREG-0612, (June 28, 1985)
This GL indicated that (1) all licensees had completed the requirement to perform a review and submit a Phase I and a Phase II report, (2) based on the improvements in heavy loads handling obtained from implementation of NUREG-0612 (Phase I), further action was not required to reduce the risks associated with the handling of heavy loads, (3) a cost-benefit analysis of PWR polar crane conversion to single-failure-proof was not cost beneficial, and (4) a detailed Phase II review of heavy loads was not necessary and that Phase II was considered completed.
Bulletin 96-02, Movement of Heavy Loads Over Spent Fuel in the Reactor Core, or Over Safety-Related Equipment (April 1996)
This bulletin was initiated because of load drop analysis performed by the Oyster Creek nuclear power plant. The bulletin: (1) alerted licensees to the importance of complying with existing regulatory guidelines on the control and handling of heavy loads, (2) reminded licensees of their responsibilities for providing adequate protection of public health and safety when handling heavy loads during plant operation, and (3) alerted licensees to the potentially high consequences that may result from a cask drop, and the importance of taking measures to mitigate such consequences in addition to measures to preclude the load drops.
This bulletin required licensees to:
- report within 30 days of the date of the bulletin, indicating the review of their plans and capabilities to handle heavy loads while the reactor is at power (in all modes other than cold shutdown, refueling, and defueled) in accordance with existing regulations, and
- provide a statement of the capability of performing the actions necessary for safe shutdown in the presence of radiological source term that may result from a breach of the dry storage cask, damage to the fuel, and damage to safety-related equipment as a result of a load drop inside the facility.
Generic safety issue proposed by NRR, Potential Risk and Consequences of Heavy Load Drops in Nuclear Power Plants (April 1999)
NRR had previously studied the issue as part of the Dry Cask Storage Action Plan, and later as the Heavy Load Control (HLC) and Crane Issues Task Action Plan prior to requesting assistance from RES.
2 CRANE OPERATING EXPERIENCE AT U.S. NUCLEAR POWER PLANTS The entire Nuclear Documents System (NUDOCS) database was searched for documents relating to cranes for the period 1968 through 1999. Additional documents were also obtained from industry, through licensee event reports, and other public documents available on the Internet. Given the time period, crane events recorded included those occurring during
6 construction and operation, and in some instances, during decommissioning. Each crane related document was reviewed and critical information was entered into a database for further analysis. These issues are listed in Appendix A, Crane Events at U.S. Nuclear Power Plants 1968 through 1999.
2.1 Crane Event Database Categories and Subcategories To analyze crane issues, several general categories were established, most with several subcategories. Once this information was input, sorts were performed to look for trends and patterns.
Table 1: Crane event database categories and subcategories General Event Category Event Subcategories Plant and event date Docket, plant name, event year, event month, type, operating/post NUREG-0612 Crane type Reactor building, polar, auxiliary, refueling/manipulator, spent fuel pool, tower, mobile, other Crane component deficiency Structure, control, brakes, rails, fasteners, unknown, none Reported administrative cause for event Not following procedures, poor procedures, test performance, load path inadequacy, ventilation inadequacy, maintenance, engineering, operations, unknown, none Safety Implication of event Death, injury, radiation release, load slip, load drop (below the hook),
crane component drop (above the hook), equipment damage, loss or partial loss of power, none Load description for slip or drop events Load/event description, height of drop or slip 2.2 Analysis of Documented Crane Issues A review of crane documents in NUDOCS for the period 1968 through 1999 resulted in approximately 300 different issues. Depending on the severity of each issue, each issue may be discussed in several documents. Most are administrative (not following a procedure, load path issues, noncompliance with technical specifications, inadequate crane operational testing prior to use, etc.) and few relate to problems encountered when lifting loads of approximately 30 tons or more. The following figures not only include a wide span of operating experience, but also include a wide variety of crane types, some of which are not used at operating nuclear facilities today. Figures 1 through 11 present nuclear crane operating experience as a whole regardless of the weight of the load being lifted, or whether the lift was done during construction, during an outage, or during plant operation. Section 3.0 discusses a subset of information contained in this section in that it contains an analysis of crane operating experience at nuclear power plants that have an operating license, and only for those loads that are classified in this report as very heavy (greater than approximately 30 tons).
7 2.2.1 Reported Crane Issues Figure 1, Documented crane issues at U.S. nuclear power plants (1968-1999), shows the total number of reported crane issues in two year increments. Crane issues were reported by individual licensees, through NRC documents and inspection reports, by vendors, and the public. The crane issues shown occurred during both construction and operation. Issues involving small hoists, light loads, or rigging issues are generally not included. Figure 1 also shows the total number of nuclear power plants that were licensed to operate during each of the two-year time periods.
As shown by the figure, there has been an increase in the number of issues over the last decade when compared to the first two decades. For example, the average number of issues per licensed plant per calendar year [i.e., (number of issues)/(average number of operating plants) divided by the number of calendar years in the sample] was (1) 0.10 for the period 1968 through 1979, (2) 0.10 for the period 1980 through 1989, and (3) 0.15 for the period 1990 through 1999. Although there was an increase in the number of reported crane issues in the last decade, the severity of crane issues as measured by the number of drops, slips, deaths or injuries was not proportional to the increase in the number of operating nuclear power plants (see figures 8 and 10).
8 Figure 1: Documented crane issues at U.S. nuclear power plants (1968-1999) 2.2.2 Crane Reports Due to Not Following Procedures Figure 2, Crane issues caused by not following procedures (1968-1999), shows the percentage of crane issue reports that were caused by not following specific crane operating procedures.
As shown in the figure, the percentage of crane issue reports caused by not following specific crane operating procedures has been cyclic, with an overall average of approximately 36 percent. Section 2.2.5 also discusses other forms of not following procedures such as not maintaining proper system alignment, room ventilation issues, load path requirements, poor procedures, etc.
9 Figure 2: Crane issues caused by not following procedures (1968-1999) 2.2.3 Crane Issue Distribution by Crane Type For the 297 reported crane issues during the period 1968 through 1999, Figure 3, Crane issue distribution by crane type (1968-1999), shows the distribution by crane type. The number of crane issues for each crane type was not broken down by reactor type, operational phase, or weight of load at the time of the event. Crane types include polar, spent fuel pool (SFP), tower, auxiliary, refueling/manipulator (RF/MC), reactor (RX) building, mobile, and other. The category other refers to cranes which do not specifically fit into one of the previous categories, and could include turbine building cranes, special cask handling cranes, unspecified cranes, or miscellaneous cranes used inside or outside of areas containing safety-related components.
The figure does not represent failure rates of different types of cranes, only the total number of documented crane issues. For example, in general, there would be many more lifts performed using the spent fuel pool or refueling cranes than mobile cranes, so a greater number of issues would be expected for spent fuel pool cranes. In addition, there are approximately twice as many PWRs (generally have polar cranes) than BWRs (generally having reactor building cranes, which are parallel track overhead bridge cranes), consequently, a greater number of issues would be expected for polar cranes than for reactor building cranes.
10 Figure 3: Crane issue distribution by crane type 2.2.4 Crane Issues Due to Hardware Deficiencies While reviewing the documented crane issues, several hardware issue categories evolved including; Unknown, Brakes, Fasteners, Rail, Components, Control Systems, Structure, and None. The crane issue was assigned one of the following categories:
Unknown. A crane malfunction had clearly occurred, but that the document did not list the component that caused the event.
Brakes. Includes malfunctions, design errors.
Fasteners. Includes loose bolting, failed fasteners, design errors.
Rail. Includes rail failures, out of alignment issues, design errors.
11 Components. Miscellaneous crane component deficiencies or malfunctions involving components other than brakes, fasteners, rails, control systems, or structures.
Control Systems. Miscellaneous control system deficiencies involving malfunctions, or design errors.
None. No crane hardware issues or deficiencies exist.
As shown in Figure 4, Crane issues due to hardware deficiencies (1968-1999), of the 297 crane issues, 118 involved actual equipment or hardware problems. The remainder (179), did not involve hardware deficiencies and were categorized as None.
2.2.5 Crane Issues Caused by Programmatic Issues Upon review of the 297 crane issues, a cause of the issue was either listed in the crane issue report or was determined by the available facts presented in the document. The crane issue was assigned one of the following categories:
Operations. Operations department failed to provide the proper conditions for load movements.
None. Unable to conclude that any programmatic deficiency existed.
Poor Procedure. Procedure was followed, but was insufficient.
Engineering/Design. Incorrect design, modification, or test parameters specified by engineering.
Maintenance. Generally related to poor maintenance repair activities.
Load Path. Crane failed to travel the correct safe load path.
Didnt Test. Failure to perform crane surveillance or operating procedures.
Ventilation. Failed to establish proper room ventilation prior to fuel movement.
Unknown. Insufficient information to conclude the type of programmatic issue.
Figure 5, Crane Program Implementation issues (1968-1999), shows the distribution of causes for the documented crane issues.
12 Figure 4: Crane issues due to hardware deficiencies Figure 5:
Crane Program Implementati on issues (1968-1999) 2.2.6 Safety Implication of Crane Issues Several outcome categories were established for each crane issue or event:
13 Load Slip. See Section 1.2.
Crane Component Drop. Any crane components located above the hook that may have separated from the crane and fallen.
Loss or Partial Loss (P-Loss) of Power. Refers to a condition where one or more incoming electrical lines lose power.
Death. May by one or more deaths per single event.
Load Drop. See Section 1.2 Injury. May be one or more injuries per single event.
Equipment Issue. Some equipment not per design, is defective, has failed, or has been damaged.
None. No impact on plant equipment, workers or the public.
Since a load drop could also result in damage to plant equipment, more than one category could be affected. Consequently, one crane event may result in more than one piece of equipment being damaged, or more than one death or injury. Of the 297 crane issues, the total number of outcomes total 324. Figure 6, Number of crane issues by category (1968-1999) indicates the number of crane issues or events for each category, and not the quantity of items affected for each category.
Figure 6:
Number of crane issues by category (1968-1999) 2.2.7 Crane Type Involved in Load Slip or Drop
14 During the period 1968-1999, there were 14 reported events involving a crane load drop, and 4 involving a load slip. Figure 7, Crane type involved in load slip or drop (1968-1999), shows the crane type involved in load slips and load drops. In addition, Table 2, Reported crane issues involving a load drop or a load slip, provides description of each event.
Load Drops:
As shown in Figure 7 and Table 2, there have been 14 reported load drop events for the period 1968 through 1999. Of these, four involved very heavy loads (greater than approximately 30 tons). Three of the four very heavy load drop events occurred during construction, while the fourth event occurred at an operating facility. The operating event was minor and resulted in the bending of two reactor head alignment pins when the crane operator inadvertently lowered the head too far (i.e., crane control functions were not lost).
Load Slips:
As shown in Figure 7 and Table 2, there have been four reported load slip events for the period 1968 through 1999. Three of these load slips involved very heavy loads at operating facilities.
The Dresden event was not risk significant, since no equipment was impacted when the reactor vessel head slipped lower when attempting to raise the head. The Comanche Peak event could have caused significant damage had the reactor coolant pump motor continued in its descent, impacting the reactor coolant piping. The Arkansas event was similar to the Dresden event and was not risk significant.
Figure 7:
Crane type involved in load slip or drop (1968-1999)
15 Table 2: Reported crane issues involving a load drop or a load slip Plant (Type)
Event Date Event Type Licensing Status During Event Very Heavy Load Crane Event Description Ginna (PWR)
July 1969 Load Drop Under construction Pre-NUREG-0612 Yes Other An assembly was dropped (due to a crane brake failure) which included the core barrel, the thermal shield, lower core plate and attached internals weighing about 90 tons. The assembly was partially supported during its fall by the crane brake. The assembly tilted slightly as it fell approximately six feet to a temporary storage support which acted as an energy absorber. Evaluation of the event indicated that the crane motor overheated, the electromagnetic brake failed and a backup mechanical brake was removed as part of a modification by Westinghouse.
Palisades (PWR)
Sept.
1970 Load Drop Operating License Pre-NUREG-0612 No Auxiliary A cable on a 25 ton auxiliary crane broke during a transfer of a control rod drive mechanism (CRDM) support tube from the reactor vessel head area to a disassembly area inside containment. The broken cable allowed the CRDM support tube, including the crane block and hook to fall approximately 22 feet to the reactor vessel head. The crane operator bypassed the upper limit electrical interlock and drove the crane sheave into the mechanical stop, breaking the crane cable. Visual damage appear to be limited to gouges on the flange surfaces of two CRDM housings, and bending of the dropped support tube.
Indian Point 3 (PWR)
Jan.
1971 Load Drop Under Construction Pre-NUREG-0612 Yes Tower The reactor vessel underwent an unscheduled descent while it was being hoisted prior to its placement. It was not clear what caused the descent. Two failures occurred, (1) the crane cable, and (2) the pinion gear bracket to base plate welds on the hoist mechanism itself. The order of the failures was not known. The time of the descent was certified to be between 15 and 60 seconds. It was concluded that no damage to the pressure vessel occurred as a result of the incident.
Fermi 1 (Sodium cooled fast reactor)
Oct.
1972 Load Drop Shutdown No Other While transferring fuel from an auxiliary fuel storage facility to the Fuel and Repair Building, a crane operator inadvertently actuated the raise instead of the lower control, causing the 1/4" bolt in the shackle holding the subassembly to fail. As a result, the subassembly fell 27 feet into the transfer tank.
Table 2: Reported crane issues involving a load drop or load slip (continued)
Plant (Type)
Event Date Event Type Licensing Status During Event Very Heavy Load Crane Event Description 16 Dresden 2,3 (BWR)
May 1976 Load Slip Operating License Pre-NUREG-0612 Yes Reactor Building The reactor building crane was being used to reinstall the Unit 2 reactor vessel head, using an inching motor. At one point, upon termination of downward drive, the head dropped abruptly approximately 15 inches before the brake engaged. A second abrupt drop was observed before the head was seated on the reactor. Both drops occurred as the head was being guided down over the reactor vessel studs, with thread protectors installed on four studs being used as guides. No forcible contact with the flange or studs occurred, and no damage resulted to either the crane or reactor components.
Troubleshooting of the brake discovered sporadic arcing of new contacts at the time of inching motor drive termination. The inching motor portion of the recent modification as tagged out of service.
River Bend (BWR)
Mar.
1983 Load Drop Under Construction Pre NUREG-0612 Yes Tower A 400 ton form assembly for the containment shield building roof was being lifted to the top of the cylindrical containment shield building, after which concrete would have been poured to form the shield roof. The day before, the 1.5 inch thick steel containment building dome had been successfully lifted and placed on the containment building by the same crane. When the form was about 30 feet above its assembly area and was about to be moved to position for lifting and placement on the shield building, the crane mast buckled and the shield form fell to the ground and the crane collapsed. Except for the shield form, no permanent structures or equipment were damaged. Cause of the crane failure was not determined.
Turkey Point 4
(PWR)
April 1983 Load Drop Operating License Post NUREG-0612 No Spent Fuel Pool An irradiated fuel assembly was being lifted out of a spent fuel storage rack using the spent fuel pool bridge crane. The two limit switches on the crane failed to stop the upward travel and the lift continued until the lifting tool jammed against the crane. The cable parted and the fuel assembly and lifting tool then fell back into the storage rack.
Three Mile Island 2 (PWR)
Dec.
1985 Load Drop Shutdown No Other While loading fuel assembly end fittings into a defueling canister, an end fitting became stuck in the canister. During attempts to reposition the stuck end fitting with the one ton jib crane, the defueling canister and support sleeve were dislodged from the canister positioning system, and dropped. The canister and sleeve fell approximately 1-1/2 feet onto the top of the debris bed in the reactor vessel. The dropped load weight was 2200 pounds, while the crane was rated at 2000 pounds.
Table 2: Reported crane issues involving a load drop or load slip (continued)
Plant (Type)
Event Date Event Type Licensing Status During Event Very Heavy Load Crane Event Description 17 Quad Cities 1 (BWR)
Sept.
1989 Load Drop Operating License Post NUREG-0612 No Spent Fuel Pool During the transfer of new fuel from the new fuel storage vault to the fuel pool, a fuel assembly was released from the refueling grapple and fell upon the spent fuel racks.
The grapple control switch was left in the release position when it was decided to lift the fuel to reposition it. The fuel was released, falling to the rack. The dropped fuel assembly and the irradiated fuel it fell on were visually examined in place from the bridge and the floor for signs of fuel damage. No damage was observed. Although no apparent damage resulted the fuel, 12 of the 32 potentially impacted fuel assemblies were discharged instead of reloaded for use in the next fuel cycle. The dropped fuel bundle was to be returned to GE.
North Anna 1 (PWR)
Feb.
1990 Load Drop Operating License Post NUREG-0612 No Spent Fuel Pool While the fuel building ventilation system was not aligned to discharge through the auxiliary building HEPA filter and charcoal absorber assembly, one fuel rod inadvertently slipped from the fuel rod handling tool due to a mechanical failure of the gripper mechanism, and dropped into its proper storage location in an uncontrolled manner. The height of the drop was not recorded, but no damage was recorded.
Fort Calhoun (PWR)
April 1990 Load Drop Operating License Post NUREG-0612 Yes Polar During the replacement of the reactor vessel head, it was inadvertently lowered too far.
It contacted the head alignment pins, bending the pins and causing superficial damage to the head flange. Prior to contacting the alignment pins, it was swinging back and forth approximately 6 inches. After contacting the pins, the head apparently dropped 6-12 inches and began pivoting on one alignment pin in a 6 foot arc. One pin was bent 12-14 degrees while the other was bent approximately 5 degrees.
Sequoyah 1 (PWR)
June 1993 Load Drop Operating License Post NUREG-0612 No Manipulator During fuel loading activities using the manipulator crane, an assembly was released prematurely, tilted over and came to rest against the south core baffle plate leaning at an angle of approximately 18 degrees from vertical. A phase A isolation, auxiliary building insolation, and containment ventilation isolation were manually initiated in accordance with procedures. No damage was done.
Peach Bottom 2
(BWR)
Sept.
1993 Load Drop Operating License Post NUREG-0612 No Auxiliary An empty irradiated component shipping liner was suspended from an auxiliary hook of the reactor building crane via an adapter about seven feet below the surface of the spent fuel pool. It dropped approximately 20 feet into the cask storage area. The adapter hook was equipped with a safety latch designed to prevent the load from slipping off the hook. The safety latch had been taped back prior to being attached to the liner sling to facilitate removal of the hook from the sling.
Table 2: Reported crane issues involving a load drop or load slip (continued)
Plant (Type)
Event Date Event Type Licensing Status During Event Very Heavy Load Crane Event Description 18 Arkansas Nuclear 1 (PWR)
Sept.
1993 Load Slip Operating License Post NUREG-0612 Yes Polar During the lift of a reactor vessel head, the polar cranes main hoist vertical motion was stopped and the head was trolleyed horizontally in the refueling canal. When the lift was resumed, the main hoist motor could not reestablish vertical motion. Subsequent attempts were made to reestablish vertical lift; but during each attempt, the head lowered slowly instead of rising.
Susquehanna (BWR)
April 1997 Load Drop Operating License Post NUREG-0612 No Auxiliary While transporting a 4000 pound toolbox using an auxiliary hoist on the reactor building crane, a nylon sling separated. One end of the box dropped approximately eight feet striking the edge of a stored Unit 2 cavity shield plug. Routine testing of slings was found to be a weakness.
Palo Verde 1 (PWR)
Feb.
1998 Load Drop Operating License Post NUREG-0612 No Other New fuel receipt inspection activities were being conducted in the Unit 1 fuel building.
The shipping container had been unbolted and a lifting rig attached. The entire container was accidently lifted approximately 2" above the platform instead of just the lid. When this condition was realized, the decision was made to lower the container, when the lid separated and the fuel was dropped to the floor. No damage was done to the new fuel.
Grand Gulf (BWR)
May 1998 Load Slip Operating License Post NUREG-0612 No Polar A core shroud tool ring became dislodged from the strong back being used to lift the ring during a planned heavy lift to remove the ring from the reactor vessel. The ring became dislodged when operations personnel changed a system alignment so that a large volume of air rose from the reactor core. When the volume of air struck the ring and lifting rig, they shook violently, resulting in two adjacent suspension points becoming dislodged (There were four total suspension points.) The ring was bearing against the top of the drywell flange, the drywell manway covers, and the drywell head studs.
Review and evaluation of the lifting rig and photographs provided no information as to why the rig failed.
Comanche Peak 1 (PWR)
Oct.
1999 Load Slip Operating License Post NUREG-0612 Yes Auxiliary During the removal of reactor coolant pump motor 1-03, the electric hoist/chain fall failed.
The 45 ton hoist was attached to the polar crane. When the hoist failed, the reactor coolant pump motor dropped approximately 15-20 feet in an unplanned descent before the hoist chain caught and prevented the motor from striking any plant structures or components. The hoist failed due to fatigue cracking of the spindle unit gear teeth.
During testing prior to its use, the hoist malfunctioned. After several attempts at performing the test, the hoist began to function properly and the job proceeded.
Improper assembly of the hoist following an overall was considered the root cause of failure.
19 2.2.8 Distribution of Load Slips and Drops (1968-1999)
Figure 8, Distribution of load slips and drops (1968-1999), shows both the number of documented load slips and load drops, and a plot showing the number of licensed power plants for the period 1968 through 1999. The crane events shown occurred during both construction and operation. Events involving small hoists, light loads, or rigging issues are generally not included. As shown by the figure, there has been a slight increase in the combined number of both load slips and load drops over the last decade when compared to the first two decades, however, this increase is substantially offset by the increase in the number of licensed nuclear power plants from 1968 to 1999.
For example, the average combined number of drop and slip events per licensed plant per calendar year [i.e., (number of events)/(average number of operating plants) divided by the number of calendar years in the sample] was (1) 0.014 for the period 1968 though 1979 (11 years), (2) 0.005 for the period 1980 through 1989 (10 years), and (3) 0.008 for the period 1990 through 1999 (10 years).
Potential reasons for the crane performance improvement shown in Figure 8 could include; (1) general implementation of lessons learned, (2) a heightened awareness of safety at operating plants as opposed to construction sites, (3) implementation of NUREG-0612 Section 5.1.1 guidelines for control of heavy loads which was issued in 1980, and (4) crane upgrades performed since initial plant licensing. Figure 8 shows a slight increase in the number of crane events is evident during the mid-1990s. Similarly, Figure 10, Distribution of crane related deaths and injuries (1968-1999) shows a comparable trend, with a concentration of nuclear crane related deaths and injuries between 1976 and 1981, with no deaths reported since 1985.
In addition, Figure 1 in Appendix C, Independent Oversight Special Study of Hoisting and Rigging Incidents Within the Department of Energy, shows an increase in the frequency of hoisting and rigging events beginning in the second quarter of 1994, although improvements were realized in the latter part of 1995.
20 Figure 8: Distribution of load slips and drops (1968-1999) 2.2.9 Crane Events Resulting in Deaths or Injuries Figure 9, Crane types involved in deaths or injuries (1968-1999), shows the number of events that led to either a death, an injury, or both a death and an injury. In reviewing deaths and injuries caused by crane operation, each event was sorted by crane type. Crane types were put into eight different categories.
Tower: Consists of a vertical tower and either a fixed or movable jib. Generally used during initial construction.
Mobile: Movable crane having various arrangements of fixed or telescoping booms or jibs. Generally used during both construction and maintenance activities.
Other: Any of several cranes not fitting into other categories (i.e., turbine building, fuel storage cask, fuel building, radwaste building, or other cranes not specifically identified by type).
Polar: Large capacity overhead crane that operates on a circular runway, normally located inside of the containment building Refueling/Manipulator: Low capacity bridge crane used during defueling and refueling operations.
Reactor Building: Large capacity overhead crane operating on a parallel runway.
21 Spent Fuel Pool: Various types of bridge cranes. Used for moving spent fuel from one location to another.
Auxiliary: Any of several lower capacity cranes or hoists.
As shown by Figure 9, most deaths and injuries occurred while using cranes that dont lift heavy loads near safety-related equipment (i.e., tower, mobile, or other categories). These types of cranes have typically not been as well controlled and maintained in the past as are polar, reactor building, or spent fuel pool cranes.
Figure 9: Crane types involved in deaths or injuries (1968-1999) 2.2.10 Description and Distribution of Crane Related Deaths and Injuries Figure 10, Distribution of crane related deaths and injuries (1968-1999), shows the number of crane related events resulting in a death, an injury, or both a death and an injury. In some instances, more than one death or injury occurred as a result of an event. During this time period, there have been seven reported crane events that have resulted in deaths, three that have involved both deaths and injuries, and 11 events that have resulted in injuries. The highest concentration of crane related deaths and injuries at nuclear power plants occurred between 1976 and 1981. The last death in a crane related accident at a U.S. nuclear power plant occurred in 1985. For comparative purposes, Figure 10 also shows the cumulative number of nuclear power plants that had an operating license during the period from 1968 through 1999.
22 Figure 10: Distribution of crane related deaths and injuries (1968-1999)
Table 3, Deaths associated with crane operation, provides information for each death event that was associated with a crane. Of the 10 events, the event at Turkey Point 4 in March 1970 was the only one that involved a very heavy load. This event also occurred during construction, and prior to the issuance of NUREG-0612 which provided guidelines for control of heavy loads. A review of each event description indicates that most of the events resulting in deaths were caused by human error and not through crane design or material deficiencies.
23 Table 3: Deaths associated with crane operation Plant (Type)
Event Date Licensing Status During Event Very Heavy Load Crane or Rigging Failure Event Description Turkey Point 4
(PWR)
March 1970 During Construction Pre-NUREG-0612 Yes Yes The main generator stator for Unit 4, which was to be installed in Unit 3, dropped one to two feet when two vertical crane support cables snapped during a lifting operation.
The support columns for the portable crane also collapsed. One section of the support columns struck and killed an engineer.
Other falling sections injured two other personnel. Some turbine piping was damaged but no nuclear components were affected.
Haddam Neck (PWR)
December 1973 Operating License Pre-NUREG-0612 No No A worker died following a 10 feet fall while effecting repair to an overhead yard crane.
Peach Bottom 2,3 (BWR)
May 1976 Operating License Pre-NUREG-0612 No No A contractor employee fell 50 feet to his death while riding a crane hook in the radwaste building.
Comanche Peak 1,2 (PWR)
May 1976 During Construction Pre-NUREG-0612 No No Failure of a portable crane boom resulted in the deaths of two construction employees when the crane became unbalanced and the boom and a occupied personnel bucket fell to the turbine mat area.
Nine Mile Point 2 (BWR)
February 1978 During Construction Pre-NUREG-0612 No Yes Two workers were killed when a section of installed reinforcing bars collapsed when struck by a bundle of reinforcing bars being handled by a crane.
Perry 1,2 (BWR)
October 1979 During Construction Pre-NUREG-0612 No No A worker was killed when he touched a crane which was in contact with a high voltage overhead line.
Marble Hill 1,2 (PWR)
February 1980 During Construction Pre-NUREG-0612 No No A worker was killed when a mobile crane got stuck in the mud and tipped over while the operator was raising the load to try to free the crane.
Byron 2 (PWR)
August 1980 During Construction Post-NUREG-0612 No No A worker was killed when he was caught between a crane counterweight and the engine housing.
McGuire 2 (PWR)
February 1985 Operating License Post-NUREG-0612 No No An equipment operator was killed when he attempted to step onto a moving manipulator crane and fell back and lodged his head between the crane and an electrical lighting panel.
Browns Ferry 2
(BWR)
March 1985 Operating License Post-NUREG-0612 No Yes A maintenance worker was killed and three others were injured when they were struck by a falling crane hook inside the unit 2 turbine building. The accident occurred when the overhead crane cable parted. The 25-ton capacity hook dropped through the roof of a temporary building where the maintenance workers were located.
24 2.2.11 Distribution of Crane Issues by Facility Figure 11, Distribution of crane issues by facility, on a per unit basis (1968-1999), shows the number of crane issues documented against each nuclear power plant facility, divided by the number of units (i.e., units that received an operating license, or were substantially completed) at that facility. Since there are many facilities that had units canceled, judgement was used in determining how many plants were substantially completed, but did not receive an operating license. Four nuclear facilities reported no crane events; Hope Creek, Kewaunee, Waterford, and Watts Bar.
Figure 11: Distribution of crane issues by facility, on a per unit basis (1968-1999)
25 3
LICENSEE CRANE OPERATING EXPERIENCE AT OPERATING NUCLEAR POWER PLANT FACILITIES (POST NUREG-0612)
At an initial meeting in May 2000, the Reactor Generic Issue Review Panel decided that the generic issue scope should be limited to (1) loads of approximately 30 tons or greater (designated as very heavy), and (2) commercial operating nuclear power plants. A representative sample of crane operating experience was obtained from nine nuclear power plant facilities consisting of 19 individual power plants. This data was put into a database, and based on the sample, estimations of the number of very heavy loads lifts was made. Crane issues or events information obtained from searching NUDOCS files, licensee event reports, other licensee documents, and industry documents was used to form the basis for the industry operating experience.
3.1 Pilot Plants for Crane Program and Operating Experience Reviews Since many hardware and programmatic changes took place with the advent of NUREG-0612 in 1980, it was determined that this crane study should include only crane operational experience since that time. From January 1980 through October 1999, U.S. nuclear power plants have operated for a combined time of approximately 1920 years. The combined operational period for the nine facilities that were visited was approximately 276 years, which is approximately a 14 percent sample. The crane operating experience sample included plants of varying designs and ages. Most were multi-unit facilities, allowing more lift data to be retrieved.
Table 4, Pilot plants for crane program and operational experience reviews, lists the facilities visited.
26 Table 4: Pilot plants for crane program and operational experience reviews Plant Design Type MWt Commercial Operation Date Onsite Visit Date Browns Ferry Units 1,2,3 BWR-Mark 1, GE 4, (AE) TVA 3293 3293 3293 1974 1975 1977 9/14-9/15/2000 Comanche Peak Units 1,2 PWR-Dry ambient, Westinghouse 4 Loop, (AE) Gibbs and Hill 3411 3411 1990 1993 11/27-11/29/2000 Diablo Canyon Units 1,2 PWR-Dry ambient, Westinghouse 4 Loop, (AE) PG&E 3411 3411 1985 1986 9/21-9/22/2000 Dresden Units 2,3 BWR-Mark 1, GE 3, (AE) S&L 2527 2527 1970 1971 7/11-7/13/2001 Grand Gulf BWR-Mark 3, GE 6, (AE) Bechtel 3833 1985 12/11-12/13/2000 Limerick Units 1,2 BWR-Mark 2, GE4, (AE) Bechtel 3458 3458 1986 1990 12/4-12/5/2000 Oconee Units 1,2,3 PWR-Dry ambient, B&W, (AE) Bechtel 2568 2568 2568 1973 1974 1974 9/27-9/28/2000 Oyster Creek BWR-Mark 1, GE 2, (AE) Brown and Root 1930 1969 8/21-8/22/2000 Palo Verde Units 1,2,3 PWR-Dry ambient, CE80, (AE) Bechtel 3800 3876 3876 1986 1986 1988 11/15-11/17/2000 3.2 Crane Operating Experience at Pilot Plants (Post NUREG-0612)
Table 5, Total number of lifts with loads of approximately 30 tons or greater, lists post NUREG-0612 (1980) crane lift data obtained from nine operating pilot facilities. The data was retrieved from the pilot plants were obtained through actual searches of crane lift records, and/or by reviewing the typical number of lifts performed during routine outages and special outages.
Items lifted include both safety and nonsafety related components. The total number of very heavy load lifts for the nine pilot facilities was approximately 7600.
27 Table 5: Total number of lifts with loads of approximately 30 tons or greater Facility Number of very heavy load lifts Browns Ferry 1,2,3 980 Comanche Peak 1,2 230 Diablo Canyon 1,2 344 Dresden 2,3 554 Grand Gulf 118 Limerick 1,2 950 Oconee 1,2,3 1656 Oyster Creek 504 Palo Verde 1,2,3 2277 3.3 Estimated Crane Operating Experience at US Nuclear Power Plants (Post NUREG-0612)
To estimate the total number of lifts greater than approximately 30 tons (designated as very heavy for this report) for all US nuclear power plants, it was necessary to normalize Table 5 lift data, taking into consideration how many refueling cycles had occurred, and the design type of the plant. The number of lifts per refueling cycle for each design type was then used to estimate the number of lifts occurring at the similar remaining power plants. The total number of estimated very heavy load lifts for all US nuclear power plants that operated from 1980 through October 1999 was approximately 47400.
3.4 Very Heavy Load Slips and Drops at Operating Facilities (Post NUREG-0612)
Of the estimated 47400 lifts, there were two load slips and one load drop that involved very heavy loads. A load slip is defined as a situation where the load may descend uncontrollably, but come to a stop without impacting or damaging other equipment. A load drop is defined as a situation where the load may descend uncontrollably, but impacts other equipment and does damage. Table 6, Very Heavy Load slips and drops occurring at operating nuclear facilities (Post NUREG-0612) provides information on one load drop event and two load slip events. Of the two very heavy load slip events (Arkansas Nuclear One-1, and Comanche Peak 1), both were caused by crane deficiencies. The reactor head load drop event at Fort Calhoun was not caused by crane deficiencies, but by operator error. The load did not totally drop, but did impact the reactor head alignment pins while lowering the reactor head, and was conservatively classified in this report as a dropped load. The dropped load did not result in a radiation release; or risk to licensee personnel or the public.
28 Table 6: Load* slips and drops of very heavy loads occurring at operating nuclear facilities (Post NUREG-0612)
Plant Event Date Load Slip Load Drop Fort Calhoun (PWR)
April 1990 While lowering the reactor head, it cocked slightly, catching on alignment pins, bending two.
(This event was caused by operator error.)
September 1993 When removing the reactor head, the head was trolleyed horizontally.
When a vertical lift was attempted, the head instead lowered. (This event was caused by a component deficiency.)
Comanche Peak 1 (PWR)
October 1999 A gearbox in an auxiliary hoist (attached to the polar crane) failed, lowering the reactor coolant pump motor about 15-20 feet. The load came to rest before impacting any equipment. (This event was caused by a component deficiency instigated by repair activities.)
- This table only includes loads that have been classified as very heavy (approximately 30 tons or greater) 3.5 Crane Load Drop Event Tree An event tree was developed (see Figure 12, Load drop event tree) assuming that the load drop was the initiating event. Probabilities for each branch were conservatively estimated using information gathered from the Pilot Plant licensees, NUREG-0612, and WASH-1400, Reactor Safety Study. Because of the vast differences between reactor safety system layout even within the same design type [i.e., BWR vs. PWR, or nuclear steam supply system (NSSS) vendor], general statements about the potential consequences of very heavy load drops at various locations within a nuclear plant is outside the scope of this study.
From a deterministic standpoint, very heavy load drops may be more risk significant at BWR plants than PWR plants because of the location of the BWR spent fuel pool on the upper floor (refueling floor) of the reactor building, and the heavy loads that the refueling floor would experience. This situation is worsened for BWRs that have a Mark I containment which places the torus directly below the equipment hatch in the reactor building. Should a load drop occur while the load is being lowered down the equipment hatch to ground level from the refueling floor (approximately 100 feet), the torus could be punctured. Accident mitigation could be compromised given a punctured torus (emergency core cooling system pump failure) or during suppression pool cooling. A heavy load drop that would penetrate the refueling floor could also disable an isolation condenser (installed at some BWRs) which would also compromise the plants capability to cope with decay heat removal following a station blackout. Other scenarios exist where individual trains of safety-related systems could be disabled, but not to the point where system redundancy or diversity would be eliminated.
29 Movements of very heavy loads near PWR spent fuel pools generally encounter few risk-significant systems because of their location at or near ground level. Drops of loaded spent fuel casks over the spent fuel pool, on the pool wall, or in the decontamination area have been analyzed by various licensees (see Table 8, Heavy load drop calculations), showing little impact on the health and safety of the public.
Number of very heavy load lifts per reactor year The number of very heavy load lifts per reactor year (25) was calculated by taking the total number of very heavy load lifts (47400 lifts) that occurred since 1980 or commercial operation, which ever was the latest, and dividing it by the total number of reactor years for the same set of power plants having an operator license (1920 years). This value was then used as the starting point for branch event probabilities as discussed in this section.
Load Drop For very heavy loads occurring at plants having an operating license, and after the issuance of NUREG-0612, there were no actual load drops. To be conservative, one very heavy load drop was assumed to occur during the period of interest (see Table 6). Assuming that the number of very heavy load lifts was approximately 47400, the load drop frequency (drops/number of lifts) was calculated to be approximately 2E-05 (1/47400 lifts).
Drop Over Safe Shutdown Equipment (On Level)
The probability of a drop over Safe Shutdown Equipment (SSE) would be related to the probability of the failure to follow procedures. As shown in Figure 5, a large percentage of crane issues are either related to not following procedures, or not properly implementing procedures. For the purposes of this assessment, it was conservatively assumed that all crane issues were the result of the failure to follow procedures and could have caused a drop over an SSE. This would result in a probability of approximately 2E-03 failures per lift. Wash-1400 provides a human reliability estimated failure rate of 1E-02 (based on data from the United Kingdom Atomic Energy Agency and the U.S. military) was used as the upper bound for this study. Both of these values will receive additional analyses.
Safe Shutdown Equipment Needed (On Level)
This is largely unknown; however, given the lack of on-level safe shutdown equipment, and the separation of redundant trains, the probability that safe shutdown equipment would be totally compromised is very small. Crane travel interlocks would have to fail, procedures would have to be violated, operations staff would not have maintained redundancy in systems during the load lift and transfer, and load contact with the safe shutdown equipment would not be incidental. It is estimated that this probability would be between 1E-03 and 5E-03 per event.
Floor Breach Since those licensees that were visited as part of this study (see Table 7) had procedural load lift height guidance for differing load weights, and routine guidance to minimize the load lift height, a floor breach would seem very unlikely unless the crane operator blatantly failed to
30 follow established procedures. The probabilities for each of the three branches in Figure 12 that would involve a floor breach are conditional upon previous failures. The logic for the factor was the degree of crane operator error and plant operations error during the load lifts. For example, for the worst case scenario, the following sequence would already have taken place; (1) a very heavy load drop occurs, (2) the load was dropped over SSE, and (3) the SSE was needed for accident mitigation. For the upper two branches, the load either didnt drop on SSE, or if the load did drop on SSE, the SSE wasnt needed for accident mitigation. Consequently, using guidance provided in NUREG-0612, the probability for a floor breach was reduced by a factor of 10 from the worst case at 1E-01 to 5E-01 to the best case at 1E-02 to 5E-02.
Safe Shutdown Equipment Below Level Depending upon the load path, there may be SSE below the level over which the load would be transported. This could be in the form of controlling instrumentation or mechanical fluid systems. The investigative level of this study (which was cursory) did not discover situations where redundancy or diversity would be eliminated. For the purposes of this study, the probability that an SSE exists below level was conservatively assumed to range between 1E-01 and 8E-01. The higher probability value (8E-01) shown in the worst case pathway was chosen because of potential common cause failures due to other preceding failures in the same pathway.
Safe Shutdown Equipment Needed (Below Level)
Transporting very heavy loads over equipment that would be necessary for plant accident mitigation would not be a conservative practice, is once again related to judgement or performance errors on the part of the crane operator and on plant operators. NUREG-0612 estimates that the probability of failure to follow a given procedure is between 1E-02 and 5E-02.
31 Figure 12: Load drop event tree 3.6 No Accident Sequence Precursor Events Involving Cranes A review of Accident Sequence Precursor (ASP) data for the period 1985 through 1999 was performed to determine if any crane related event met the thresholds for risk significance. The ASP program identifies and categorizes precursors to potential severe core damage accident sequences. Accident sequence precursors are those that, if additional failures occurred, could have resulted in inadequate core cooling, causing severe core damage. The ASP program analyzes potential precursors and calculates their conditional core damage probability (CCDP).
The CCDP is the probability that the event or condition could have progressed to core damage given the existence of the failed or degraded protective or mitigating features or initiating event.
To be classified as an ASP event, the event must have a CCDP of at least 1.0 x 10-6.
The most potentially risk-significant crane events involved loss or partial loss of offsite power.
For the period 1985 through 1999, there were six such mobile crane events. These are summarized in Table 7.
32 Table 7: Crane events resulting in a loss or partial loss of offsite power Plant Event Date Description Peach Bottom 2 (BWR)
August 1987 While Unit 2 had been shutdown for five months, an 80 ton mobile crane contacted an energized 220 KV line resulting in tripping of the Unit 2 startup source line. Both Units 2 and 3 were affected. Unit 3 C RHR was restored within 10 minutes. Unit 2 C RWCU pump was restored within 37 minutes, and RHR was returned to service within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. (LER 277 016)
Fermi 2 (BWR)
December 1991 While in cold shutdown, a mobile crane contacted an energized 120 KV overhead electrical line twice. The circuit opened and closed momentarily for each contact, but did not cause a loss of offsite power. (No LER was written)
Palo Verde 3 (PWR)
November 1991 While Unit 3 was in hot standby, a 35 ton mobile crane contacted a 13.8 KV overhead line causing a partial loss of offsite power. The crane was not grounded, was not level, the friction brake was not set, and the crane was left unattended when its boom rotated into the power line. (LER 530 010-01, also an augmented inspection team (AIT) inspection was performed)
Diablo Canyon 1 (PWR)
March 1991 While Unit 1 was in a refueling outage, loss of offsite power caused by mobile crane when it got too close to a 500 KV electrical line. The 230 KV startup power system had been cleared for maintenance and was not available. RHR capability was lost for less than one minute, and the spent fuel pool pumps were inoperable for approximately 23 minutes. An Unusual Event was declared. (LER 275-91-004-01, also an AIT inspection was performed)
Nine Mile Point 2 (BWR)
September 1992 While Unit 2 was at 100 percent power, a mobile crane boom got too close to one of two 115 KV lines, tripping the line and causing a partial loss of offsite power. Division I and II EDGs ran loaded for approximately 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> each. The 115 KV line was restored within approximately 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />. (LER 410-92-020)
Indian Point 3 (PWR)
March 1995 While Unit 3 was in cold shutdown, a mobile crane in the Indian Point 2 owner controlled area shorted the C phase of the 138 KV feeder to ground causing a loss of offsite power. Emergency power was provided by two EDGs. (LER 286-95-004)
Of the six crane events described in Table 7, two licensees had Augmented Inspection Team (AIT) inspections (Palo Verde and Diablo Canyon). However, none of the six mobile crane events met the minimum risk threshold requirements to be classified as an ASP event.
33 4
LICENSEE LOAD DROP CALCULATION METHODOLOGY AND RESULTS A sampling of load drop calculations obtained from each facility that was visited indicated that calculational methodologies and assumptions varied from licensee to licensee, producing different end results. Heights of load drops, plant locations for postulated load drops, contact area at impact, materials property values, and weights of loads varied greatly. The Oyster Creek calculation for a drop of a 45 ton fuel cask over a reinforced concrete 16 inch thick slab was the most restrictive, with an allowable drop height of 2.77 inches. Some facilities performed load drop calculations using ballistic type equations based on high velocity and low mass situations. Each licensee used load drop calculations to determine transport height restrictions in their heavy load procedures. These restrictions would be based on conservative engineering analyses. Table 8, Load drop calculations for very heavy components, provides a sampling of load drop calculations from the facilities that were visited.
Table 8: Load drop calculations for very heavy components 34 PLANT CALC DATE LOAD WT (tons)
HT TARGET COMPOSITION STRIKING VELOCITY ASSUMPTIONS CALCULATION OUTCOMES Grand Gulf (BWR) 8/15/78 Bechtel Drywell head 61.5 5ft (air)
- Refueling floor; 9" RC slab on 3" decking (non-composite), slab supported on W36x300 beams
@ 6'4" spacing 17.9 ft/sec Used an equation for penetration of 12" diameter missiles. 100% of flange will contact the floor.
- Depth of penetration 2.8"
- 9" RC slab = 6.9
- W36x300 = 5.9 Grand Gulf (BWR) 8/17/78 Bechtel Drywell head 61.5 30 ft (air)
- Reactor well; 1.5" wide sleeve, radius of 16'-3/4" 43.9 ft/sec Drywell head hits the sleeve Drywell head crushes the sleeve, and continues downward, but doesnt compromise the integrity of the RPV Grand Gulf (BWR) 8/16/78 Bechtel RPV head 117 5 ft (air)
- Refueling floor, 4-'0 thick RC 17.9 ft/sec 100% of flange will contact the floor.
-Depth of penetration 4.4"
-For simple support, =9; for fixed support, 1 Grand Gulf (BWR) 4/4/78 Bechtel Steam separator 68 17 ft (water)
- Spent fuel pool; Steam separator area, 52" thick slab with 1/4" liner plate 21.5 ft/sec Steam separator falls in water
-Assuming a 1/4" plate, the depth of penetration
=.7" (unsatisfactory)
-Assuming a 52" concrete slab, depth of penetration = 6.2"
-Assuming an interface forcing function, depth of penetration = 2.6"
-Using a structural response and ratioing, the slab response will not exceed the acceptable ductility ratio of 10.
Grand Gulf (BWR) 7/18/78 Bechtel Steam dryer 40 23 ft (
air)
- Dryer storage area, 52" thick slab with 1/4" liner plate 38.5 ft/sec
- For the 1/4" liner plate, the equation appears to spread out the load over an entire cylinder with a diameter of 238"(same for the slab) as opposed to an annulus.
- Assuming a 1/4' plate, the depth of penetration =.09"
- Assuming a 52" thick concrete slab, depth of penetration = 5.4"
- 5.3 Oyster Creek (BWR) 10/29/99 EQE Fuel cask 45 6" (air)
- Refueling floor; At the center of beam 5B27; slab thickness 16";
beam width 36,"
beam depth 30";
various rebar 8-15,
- 8 5.7 ft/sec
- Allowable drop height = 7.01"
Table 8: Load drop calculations for very heavy components (continued)
PLANT CALC DATE LOAD WT (tons)
HT TARGET COMPOSITION STRIKING VELOCITY ASSUMPTIONS CALCULATION OUTCOMES 35 Oyster Creek (BWR) 10/26/99 EQE Fuel cask 45 (3.85")
(air)
- Refueling floor; center drop on slab 5S10; slab span N/S 23'-3" x E/W 20'-9"; slab thickness 16";
rebar #6 @7" and 18" centers, and
- 8@6, 8, &9" centers; (4.55 ft/sec)
- Allowable drop height = 3.85" Oyster Creek (BWR) 10/26/99 EQE Fuel cask 45 (2.77")
(air)
- Refueling floor; Drop on slab 5S10 adjacent to beam 5B27; slab span N/S 23'-3" x E/W 20'-9"; slab thickness 16";
rebar #6 @7" and 18" centers, and
- 8@6, 8, &9" centers (3.86 ft/sec)
- Allowable drop height = 2.77" Oyster Creek (BWR) 10/26/99 EQE Fuel cask 45 (11.58"
) (air)
- Refueling floor; Drop on slab 5S14 adjacent to beam 5B39; similar to slab 5S10 but slab thickness = 26";
(7.88 ft/sec)
- Allowable drop height = 11.58" Oyster Creek (BWR) 10/26/99 EQE Fuel cask 45 6" (air)
- Refueling floor; Drop on east wall of spent fuel pool; the wall is 6' thick and extends from the 119' level to the 72' level; 5.7 ft/sec
- Analyzed as a hard object striking a hard target; the drop would occur between columns C5 and C6 and between beam 5B21 and 5B19, and slab 5S14; target mass 10000 lb/ftsec2
- Available strain energy calculated at ~254 kipft, kinetic energy from drop ~31 kipft
- If kinetic energy of drop is set equal to the strain energy, the allowable drop height would be 49.6"
- If load is dropped directly on C6, the allowable drop height would be 49" Palo Verde (PWR) 6/4/80 Bechtel Fuel cask 125 12' (air)
- Drop from level 124.5' to the decontamination pit (~12');
27.8 ft/sec
- Assumes that the cask hits the floor exactly flat; Ductility ratio of 30 acceptable
- Thickness required to preclude spalling 71.56"; slab defection.063"; ductility ratio calculated to be 22.84
Table 8: Load drop calculations for very heavy components (continued)
PLANT CALC DATE LOAD WT (tons)
HT TARGET COMPOSITION STRIKING VELOCITY ASSUMPTIONS CALCULATION OUTCOMES 36 Palo Verde (PWR) 6/4/80 Bechtel Fuel cask 125 30' (air)
- Drop from the top of the spent fuel pool to the bottom of the cask loading pit; target slab is 7'-9" thick 44.96 ft/sec
- Ductility ratio of 30 acceptable
- Ductility ratio of 6.01 calculated, 30 is acceptable; - using a different soil reaction, ductility ratio calculated to be 9.67, 30 is acceptable Palo Verde (PWR) 6/23/80 Bechtel Fuel cask 126 1' +
rotation strike on wall
- Drop from top of spent fuel pool to the decontamination pit and then deflects to the east wall of the pit (Striking velocity on the wall =
16.133 ft/sec)
- Ductility ratio of 30 acceptable
- Calculated ductility ratio 47.09, 20 (average of beam, 10, and slab 30)
- For this situation, an energy absorbing pad was required Browns Ferry (BWR) 1/14/72 TVA Fuel cask 100 3'
- Drop on hypothetical 18" RC slab 13.9 ft/sec
- NAVDOCKS (p51)
- Cask lands flat on 16 fins, evenly distributed (4.124 ft2)
- Depth of penetration =.0892 ft Browns Ferry (BWR) 1/17/72 TVA Fuel cask 100 3'
- Drop on hypothetical 18" RC slab 13.9 ft/sec
- Compares energy absorbed to the energy the system can ultimately absorb
- energy to be absorbed = 7.2 E6, in-# energy the system can ultimately absorb = 9.35 E6 in-#
Browns Ferry (BWR) 1/18/72 TVA Fuel cask 100 3'
- Drop on 18" thick slab near supports 13.9 ft/sec
- After punching through in the area immediately adjacent to the slab support, the structural system will form two effective cantilever beams with three plastic hinges
- Punch through will occur near the column and beams in an arc, it will not go through the slab Browns Ferry (BWR) 1/27/72 TVA Fuel cask 100 6"
- Drop on 36" slab 5.675 ft/sec
- Uses a modified Petry formula for penetration
- Penetration calculated to be.015 ft
Table 8: Load drop calculations for very heavy components (continued)
PLANT CALC DATE LOAD WT (tons)
HT TARGET COMPOSITION STRIKING VELOCITY ASSUMPTIONS CALCULATION OUTCOMES 37 Limerick (BWR) 4/30/84 Bechtel (1)
Drywell head 104 3'
- Tilted drop on refueling floor, RC 24" thick, #9@8" centers (T&B);
13.9 ft/sec
- Capacity of slab based on yield-line theory, simple span, elasto-plastic design
- Doesnt appear to account for kinetic energy absorption over a small area
- Tilted drop case, strikes over 40 degrees of circumference
- Interface force = 6.35 E6 # (average=2.1 E6 #)
- Punching shear capacity appears to be high (240 psi)
- Calculated punching shear appears to be low (117psi)
- Compressive strength of concrete appears to be high
- E for concrete appears to be high
- =.8, allowable 10 ( over concrete, Zones A&B)
- 8.72, allowable 8.72 (over W36 beam, Zones A&B)
- = 7.5, allowable 8.72 (over two W36 beams, Zones A&B)
- 1.0, (over concrete, Zone C)
- 12, allowable 20 (over W24)
- = 10, allowable 12 (over two beams W24)
Limerick (BWR) 4/24/84 Bechtel (2)
Drywell head 104 3'
- Flat drop on refueling floor 13.9 ft/sec
- Drywell head lands completely flat on the refueling floor
- Interface force = 7.1 E6 #
- Flat drop case shows a greater force on the floor than does the tilted case above
- = 1.8, allowable 10 (over concrete zone A&B)
- = 1.5, 8.72 allowable 8.72 (over W36 beam, Zones A&B)
- = 1.4, (over concrete, Zone C)
- = 2, allowable 12, (over W24, zone C)
Limerick (BWR) 4/26/84 Bechtel (3)
RPV Head 92 3'
- Flat drop on refueling floor 13.9 ft/sec
- RPV head lands completely flat on the refueling floor
- Interface force = 1.23 E7 #
- = 1.8, allowable 10, (over concrete, Zones A&B)
- 3 (over W36, zones A&B)
- = 1.3, allowable 10, (over concrete, Zone C)
- 5, (over W24, zone C)
Limerick (BWR) 4/26/84 Bechtel (4)
RPV Head 92 3'
- Tilted drop on refueling floor 13.9 ft/sec
- RPV head lands tilted
- Interface force = 7.16 E6 # (average=2.39 E6 #)
- Flat drop case shows a greater force on the floor than does the tilted case above
- = 1.0 ( over concrete, Zones A&B)
- = 5.5 (over two beams, W36, zones A&B)
- 1.0, (over concrete, Zone C)
- 100, (over two beams, W24, Zone C)
Limerick (BWR) 4/26/84 Bechtel (5)
RPV Head 92 2'
- Tilted drop on refueling floor 11.38 ft/sec
- RPV head lands tilted
- Interface force = 6.55 E6 # (average=2.18 E6 #)
- ~ 20, (over W24, Zone C)
- Drop height was changed from 3' to 2' to get a lower
Table 8: Load drop calculations for very heavy components (continued)
PLANT CALC DATE LOAD WT (tons)
HT TARGET COMPOSITION STRIKING VELOCITY ASSUMPTIONS CALCULATION OUTCOMES 38 Limerick (BWR) 4/26/84 Bechtel (6)
Shield Plugs 12 3'
- Flat drop on refueling floor 13.9 ft/sec
- Flat drop calculated for over W36, 24" thick concrete, and W24
- Interface force = 3.37 E7 #
- Flat drop force for the 12 ton plugs was calculated to be greater than the tilted drop of the drywell head at 104 tons
- = 1.5, allowable 10, ( over concrete, Zones A&B)
- 1 (over W36, zones A&B)
- =1.5, allowable 10, (over concrete, Zone C)
- =2.4, allowable 10, (over W24, zone C)
Limerick (BWR) 4/26/84 Bechtel (7)
Stoplog 59 3'
- Flat drop on refueling floor 13.9 ft/sec
- Flat drop
- Contact area = 75 ft2
- Interface force = 5.3 E7 #
- = 2, allowable 10, ( over concrete, Zones A&B)
- = 1.08, allowable 8.72 (over W36, zones A&B)
- 2.5, allowable 10, (over concrete, Zone C)
- = 1.53, allowable 10, (over W24, zone C)
Limerick (BWR) 4/26/84 Bechtel (8)
Stoplog 59 3'
- Tilted drop (45 degrees) on refueling floor 13.9 ft/sec
- Tilted drop
- Contact area = 2.5 ft2
- Interface force = 1.78 E6 # (average=5.94 E5 #)
- Per an unreferenced equation, spalling of concrete will occur at a drop height of approximately 4"
- =.6, allowable 10, ( over concrete, Zones A&B)
- Punching shear capacity appears to be high (240 psi from p. 12 of calc)
- Calculated punching shear appears to be low (173 psi)
- = 4, allowable 8.72, (over W36, zones A&B)
- =.4, allowable 10, (over concrete, Zone C)
- = 100, allowable 12, (over W24, zone C)
Limerick (BWR) 4/26/84 Bechtel (9)
Stoplog 59 1'-9" 10.6 ft/sec
- Tilted drop
- Contact area = 2.12 ft
- Interface force = 1.14 E6 # (3.814 E5 #)
- ~ 20, (over concrete with embedded beams)
Table 8: Load drop calculations for very heavy components (continued)
PLANT CALC DATE LOAD WT (tons)
HT TARGET COMPOSITION STRIKING VELOCITY ASSUMPTIONS CALCULATION OUTCOMES 39 Limerick (BWR) 4/26/84 Bechtel (10)
Stoplog 38 2'
- Tilted drop (45 degrees) on refueling floor 11.35 ft/sec
- Tilted drop
- Contact area = 1.3 ft2
- Interface force = 9.41 E5 #
(average=3.134 E5 #)
- Per an unreferenced equation, spalling of concrete will occur at a drop height of approximately 7"
- 1.0, allowable 10, ( over concrete, Zones A&B)
- Punching shear capacity appears to be high (240 psi from p. 12 of calc)
- Calculated punching shear appears to be low (101 psi)
- = 1.2, allowable 8.72, (over W36, zones A&B)
- 1.0, (over concrete, Zone C)
- ~ 12, allowable 12, (over W24, zone C)
Limerick (BWR) 4/26/84 Bechtel (11)
Stoplog 38 2'
- Flat drop on refueling floor 11.35 ft/sec
- Flat drop
- Contact area = 135 ft2
- Interface force = 8.68 E6 #
- = 3.0, allowable 10, ( over concrete, Zones A&B)
- = 1.2, allowable 8.72, (over W36, zones A&B)
- Per an unreferenced equation, spalling of concrete will occur at a drop height of approximately 16.1"
- = 1.5, allowable 10, (over concrete, Zone C)
- 12, allowable 12, (over W24, zone C)
Limerick (BWR) 4/26/84 Bechtel (12)
Stoplog 38 1.5' (air) 22.5' (water)
- Flat drop back into its slot 30.9 ft/sec
- Flat drop
- assume 50% contact (831.25 in2)
- Interface force 2.73 E7 # (average=9.1 E6
- )
- Penetration based on impact duration = 1.4"
- Penetration based on missiles hitting soils =
.68" Limerick (BWR) 4/26/84 Bechtel (13)
Stoplog 38 1.83' (air) 37.75' (water)
- Flat drop into the Fuel Pool 34.5 ft/sec
- Flat drop
- Assumes 50% contact (831.25 in2)
- Interface force 3.04 E7 # (average=1.01 E7
- )
- Penetration based on impact duration = 1.7"
- Penetration based on missiles hitting soils =
.85" Limerick (BWR) 4/23/84 Bechtel (14)
Steam dryer assembly 45 6'
- Flat drop on refueling floor 19.7 ft/sec
- Flat drop
- Total contact area = 3000 in2
- Contact area for slab of interest = 1140 in2
- = 3, allowable 10, ( over concrete, Zones A&B)
- = 2.0, allowable 8.72 (over W36, zones A&B)
- =1.7, allowable 10, (over concrete, Zone C)
- = 2, allowable 12, (over W24, zone C)
Table 8: Load drop calculations for very heavy components (continued)
PLANT CALC DATE LOAD WT (tons)
HT TARGET COMPOSITION STRIKING VELOCITY ASSUMPTIONS CALCULATION OUTCOMES 40 Limerick (BWR) 4/26/84 Bechtel (15)
Steam dryer assembly 45 6'
- Tilted drop (17.46 degrees) on refueling floor 19.7 ft/sec
- Tilted drop
- Contact area = 4.06 ft2
- Interface force = 4.07 E6 # (average=1.36 E6 #)
- 1, allowable 10, ( over concrete, Zones A&B)
- = 9, allowable 8.72 (over W36, zones A&B)
Limerick (BWR) 4/26/84 Bechtel (16)
Steam dryer assembly 45 5'
- Tilted drop (14.5 degrees) on refueling floor 17.94 ft/sec
- Tilted drop
- Contact area = 4.18 ft2
- Interface force = 3.83 E6 # (average=1.28 E6 #)
- = 8, allowable 8.72 (over W36, zones A&B)
- 1, allowable 10, (over concrete, Zone C)
- = 50, allowable 12, (over W24, zone C)
Limerick (BWR) 4/26/84 Bechtel (17)
Steam dryer assembly 45 3'
- Tilted drop (8.62 degrees) on refueling floor 13.9 ft/sec
- Tilted drop
- Contact area = 7.29 ft2
- Interface force = 5.17 E6 # (average=1.72 E6 #)
- = 12, allowable 12, (over W24, zone C)
Limerick (BWR) 4/28/84 Bechtel (18)
Steam separator assembly 81.5 5'
- Flat drop on refueling floor 17.9 ft/sec
- Flat drop
- Contact area = 5.61 ft2
- Interface force = 5.12 E6 #
- = 2, allowable 10, ( over concrete, Zones A&B)
- = 2.0, allowable 8.72 (over W36, zones A&B)
- =1.8, allowable 10, (over concrete, Zone C)
- =.25, allowable 10, (over W24, zone C)
Limerick (BWR) 4/28/84 Bechtel (19)
Steam separator assembly 81.5 5'
- Tilted drop (14.5 degrees) on refueling floor 17.9 ft/sec
- Tilted drop
- Contact area = 4.97 ft2
- Interface force = 4.55 E6 # (average=1.52 E6 #)
- 1, allowable 10, ( over concrete, Zones A&B)
- = 5.5, allowable 20 (two beams, over W36, zones A&B)
- =1.5, allowable 10, (over concrete, Zone C
- = 25, allowable 20 (two beams, over W24, zone C)
Limerick (BWR) 4/28/84 Bechtel (20)
Steam separator assembly 81.5 2.5'
- Tilted drop (7.2 degrees) on refueling floor 12.7 ft/sec
- Tilted drop
- Contact area = 5.57 ft2
- Interface force = 3.595 E6 # (average=1.2 E6 #)
- = 12, allowable 20 (two beams, over W24, zone C)
Limerick (BWR) 4/28/84 Bechtel (21)
Steam separator assembly 81.5 7'
- Flat drop on refueling floor 21.2 ft/sec
- Flat drop
- Contact area on slab of interest = 7.92 ft2 weight on slab of interest = 31 tons
- Interface force = 8.56 E6 #
- = 3.5, allowable 10, ( over concrete, Zones A&B)
- = 2.8, allowable 10 (over W36, zones A&B)
Limerick (BWR) 4/28/84 Bechtel (22)
Steam separator assembly 81.5 7'
- Tilted drop (20.5 degrees) on refueling floor 21.2 ft/sec
- Tilted drop
- Contact area = 4.88 ft2
- Interface force = 5.27 E6 # (average=1.76 E6 #)
- =.7, allowable 10 (over W36, zone D)
Table 8: Load drop calculations for very heavy components (continued)
PLANT CALC DATE LOAD WT (tons)
HT TARGET COMPOSITION STRIKING VELOCITY ASSUMPTIONS CALCULATION OUTCOMES 41 Limerick (BWR) 5/8/84 Bechtel (23)
Shield plugs 85 3'
- Flat drop on refueling floor 13.9 ft/sec
- Flat drop
- Total contact area = 418.5 ft2
- Contact area on slab of interest = 181.9 ft2
- Interface force = 1.29 E8 #
- 10, allowable 10, ( over concrete, Zones A&B)
Limerick (BWR) 5/8/84 Bechtel (24)
Shield plugs 85 3'
- Tilted drop (5.3 degrees) on refueling floor 13.9 ft/sec
- Tilted drop
- Total contact area = 2.65 ft2
- Interface force = 1.88 E6 # (average=6.27 E5 #)
- 1, allowable 10, ( over concrete, Zones A&B)
- = 4.0, allowable 10 (over W36, zones A&B)
- 1, allowable 10, (over concrete, Zone C
- 12, allowable 12 (two beams, over W24, zone C)
Limerick (BWR) 5/8/84 Bechtel (25)
Shield plugs 85 2'
- Tilted drop (3.5 degrees) on refueling floor 11.3 ft/sec
- Tilted drop
- Total contact area = 2.67 ft2
- Interface force = 1..54 E6 # (average=5.1 E5 #)
- = 12, allowable 12 (two beams, over W24, zone C)
Limerick (BWR) 6/17/96 S&L Shield plug 85
- Tilted blunt drop on drywell head
- Slightly tilted drop
- Drywell head materia l thickness at impact is 1.5" SA 516 Gr 70
- Postulates the failure of two lifting lugs on the plug
- ADINA computer program used to analyze the drywell head under an increasing local load
- It is assumed that the plug rotates on a hinge (failure of a lifting lug, not the crane) so only 53 % of load hits the drywell head
- S&L doesnt provide an analysis for a sharp (small area) impact
- Area of impact = 754 in2
- Once the effective load of the plug is reduced from 170 k# to 79 k#, the strain energy to be absorbed by the drywell head was calculated to be 3402 in-kips
- Increased the capability of the head toy a DIF of 1.2, the materials can take 4774 in-kips which is about 40% higher than that caused by the plug drop
- The deflection at maximum strain energy would be approximately 8", whereas at the calculate strain energy, the drywell head will deflect approximately 5.8" Comanche Peak (PWR) 12/8/88 SWEC (4)
Reactor Coolant Pump Assembly 27.6
- 20 thick RC
- Slabs S-4 to S-8
- (General) Strain energy capacity is compared to kinetic energy of the load drop; Assumes a ductility ratio of 10; An elasto-plastic force-deflection diagram is assumed to represent the energy absorbing capacity of the structural component
- Missile area = 7 ft. Diameter
- Maximum drop height = 5" (Scabbing)
- The contact areas was changed in calculation listed as 4-1 below
Table 8: Load drop calculations for very heavy components (continued)
PLANT CALC DATE LOAD WT (tons)
HT TARGET COMPOSITION STRIKING VELOCITY ASSUMPTIONS CALCULATION OUTCOMES 42 Comanche Peak (PWR) 12/8/88 SWEC (4-1)
Reactor Coolant Pump Assembly 27.6
- 20 thick RC
- Slabs S-4 to S-8
- (General) Strain energy capacity is compared to kinetic energy of the load drop; Assumes a ductility ratio of 10; An elasto-plastic force-deflection diagram is assumed to represent the energy absorbing capacity of the structural component
- Missile area = 2 ft. Diameter
- Maximum drop height = 1' - 2" Comanche Peak (PWR) 12/8/88 SWEC (4-2)
Reactor Coolant Pump Assembly 27.6
- 26 thick RC
- Slabs S-1, 2, 3, and 9
- (General) Strain energy capacity is compared to kinetic energy of the load drop; Assumes a ductility ratio of 10; An elasto-plastic force-deflection diagram is assumed to represent the energy absorbing capacity of the structural component
- Missile area = 2 ft. Diameter
- Maximum drop height = 3' - 3""
Comanche Peak (PWR) 12/8/88 SWEC (5-1)
Reactor Coolant Pump Stator 23.8
- 20 thick RC
- Slabs S-4 to S-8
- (General) Strain energy capacity is compared to kinetic energy of the load drop; Assumes a ductility ratio of 10; An elasto-plastic force-deflection diagram is assumed to represent the energy absorbing capacity of the structural component
- Missile area = 6 ft. Diameter
- Maximum drop height = 5" Comanche Peak (PWR) 12/8/88 SWEC (5-2)
Reactor Coolant Pump Stator 23.8
- 26 thick RC
- Slabs S1, 2, 3, and 9
- (General) Strain energy capacity is compared to kinetic energy of the load drop; Assumes a ductility ratio of 10; An elasto-plastic force-deflection diagram is assumed to represent the energy absorbing capacity of the structural component
- Missile area = 6 ft. Diameter
- Maximum drop height = 1' - 3" Comanche Peak (PWR) 12/8/88 SWEC (6-1)
Reactor Coolant Pump Motor Assembly (Rotor &
Stator) 42.4
- 20 thick RC
- Slabs S-4 to S-8
- (General) Strain energy capacity is compared to kinetic energy of the load drop; Assumes a ductility ratio of 10; An elasto-plastic force-deflection diagram is assumed to represent the energy absorbing capacity of the structural component
- Missile area = 2 ft. Diameter
- Maximum drop height = 9"
Table 8: Load drop calculations for very heavy components (continued)
PLANT CALC DATE LOAD WT (tons)
HT TARGET COMPOSITION STRIKING VELOCITY ASSUMPTIONS CALCULATION OUTCOMES 43 Comanche Peak (PWR) 12/8/88 SWEC (6-2)
Reactor Coolant Pump Motor Assembly (Rotor &
Stator) 42.4
- 26 thick RC
- Slabs S1, 2, 3, and 9
- Slab S-10
- (General) Strain energy capacity is compared to kinetic energy of the load drop; Assumes a ductility ratio of 10; An elasto-plastic force-deflection diagram is assumed to represent the energy absorbing capacity of the structural component
- Missile area = 2 ft. Diameter
- Maximum drop height = 2' - 1" Comanche Peak (PWR) 12/8/88 SWEC (7/7A)
Reactor Coolant Pump Motor Assembly (Rotor &
Stator) 27.6
- 54 thick RC
- Slab S-10
- (General) Strain energy capacity is compared to kinetic energy of the load drop; Assumes a ductility ratio of 10; An elasto-plastic force-deflection diagram is assumed to represent the energy absorbing capacity of the structural component
- Missile area = 2 ft. Diameter
- Maximum drop height = 48' - 9" (Scabbing)
- Maximum drop height = 24' - 4" (to reach strain energy max)
Comanche Peak (PWR) 12/8/88 SWEC (8/8A)
Reactor Coolant Pump Rotor) 3.3
- 54 thick RC
- Slab S-10
- (General) Strain energy capacity is compared to kinetic energy of the load drop; Assumes a ductility ratio of 10; An elasto-plastic force-deflection diagram is assumed to represent the energy absorbing capacity of the structural component
- Missile area = 2 ft. Diameter
- Maximum drop height = 353' - 11" (Scabbing)
- Maximum drop height = 176' - 8" (to reach strain energy max)
Oconee (PWR) 6/1/82 (1)
- Low pressure turbine rotor 138 30 feet above turbine deck
- Turbine deck floor 11.5" thick RC
- Second floor 8" thick RC
- Base floor 48" RC
- 43.95 ft/sec at impact on turbine deck
- Methodology based on Bechtel Power Topical Report, BC-TOP-9 Rev. 2, September 1974 Design of Structures for Missile Impact
- Rotor falls with its shaft perpendicular to the floor, flat contact
- Ductility ratio of 10
- Perforation depth calculated to be 10.31," (i.e.,
the rotor will not go through the turbine deck floor)
- The drop will result in bending failure of the operating floor slab
- The second floor will be penetrated by punching shear
- The rotor will penetrate approximate 7" into the basement floor
- Will not damage any piping greater than 14.12" Oconee (PWR) 6/1/82 (2)
- Low pressure turbine rotor 138 77 feet
- 60" thick RC basement floor 70.4 ft/sec
- Methodology based on Bechtel Power Topical Report, BC-TOP-9 Rev. 2, September 1974 Design of Structures for Missile Impact
- Rotor falls down the equipment hatch
- Penetration depth of rotor = 21.12"
- Some spalling may occur
- With not prevent vital embedded systems from performing their safety related functions
Table 8: Load drop calculations for very heavy components (continued)
PLANT CALC DATE LOAD WT (tons)
HT TARGET COMPOSITION STRIKING VELOCITY ASSUMPTIONS CALCULATION OUTCOMES 44 Oconee (PWR) 10/16/75 (3)
- Spent fuel cask 24 46.5 feet (40 feet throug h
water)
- Floor of spent fuel pool 54.72 ft/sec
- Allow one trunnion or side of yoke to fail, load stabilizes, and then falls to the SFP floor
- Cask hits at approx. 11 degrees
- Uses modified Petry formula
- Penetration in steel floor plate 1.91 inches.
Actual thickness of plate on the floor is 2.25" Oconee (PWR) 5/19/89 (4)
-Spent fuel cask 100
- Floor of spent fuel pool 55 ft/sec
- Uses missile impact theory
- Very little chance of a large eccentric drop due to gaps between the cask and surrounding equipment
- Assumes that the impact is evenly distributed around the cask bottom ring
- Assumes that the cask falls through air
- Cask penetration into concrete = 11.4 Oconee (PWR) 5/19/89 (5)
-Spent fuel cask 100 46.5 ft
- Floor of spent fuel pool 46 ft/sec
- Uses missile impact theory
- Very little chance of a large eccentric drop due to gaps between the cask and surrounding equipment
- Assumes that the impact is evenly distributed around the cask bottom ring
- Assumes that the cask falls through water
- Includes buoyancy and drag effects of water
- Cask penetration into concrete = 6.8 Oconee (PWR) 5/26/89 (6)
- Spent fuel cask
- Floor of spent fuel pool
- Assumes that the largest crack possible would be 1/64" wide and could include the largest plate in the spent fuel pool (568" in perimeter)
- Assumes that 40' of water is in the pool
- The leakage rate was calculated to be 21.3 gallons per day Oconee (PWR) 11/21/80 (7)
- Spent fuel cask 29.1 27' - 9"
- Fuel rack 42.27 ft/sec
- Assumes free fall to the rack (no water)
- Assumes all the kinetic energy is absorbed in part by buoyancy force
- Actual crush tests were performed on fuel cans
- If cans are damaged, then radioactive gases are released
- 522 cells will be damaged Oconee (PWR) 12/2/80 (8)
- Spent fuel cask (TN-8) 43.4 27' - 9"
- Fuel rack 42.27 ft/sec
- Cask hits the side of the spent fuel pool
- Assumes all the kinetic energy is absorbed in part by buoyancy force
- 576 cells damaged
Table 8: Load drop calculations for very heavy components (continued)
PLANT CALC DATE LOAD WT (tons)
HT TARGET COMPOSITION STRIKING VELOCITY ASSUMPTIONS CALCULATION OUTCOMES 45 Oconee (PWR) 2/26/88 (9)
Radiologica l
consequen ces of spent fuel cask drop in pool NA NA
- Fuel rack NA
- Assumes that a maximum of 1024 assemblies damaged in the units 1 and 2 fuel pool (354 assemblies have less than 1 year decay, the remaining have 1 year decay)
-Assumes that a maximum of 825 assemblies are damaged in the unit 3 fuel pool (177 assemblies have less than 1 year decay, the remaining have 1 year decay)
- Assumes that the entire gap activity is released for the effected assembly
- No credit is given for HVAC filtration
- Beta does from plume is insignificant
- Total body dose (Rem) for units 1 and 2 =.15
- Total body dose (Rem) for unit 3 =.13
- Thyroid dose (Rem) for units 1 and 2 = 72
- Thyroid dose (Rem) for unit 3 = 72 Diablo Canyon (PWR) 9/16/86 Bechtel (1)
- RC slab, infinite thickness 8.02 ft/sec
- Assumes infinite slab thickness
- Assumes missile impact
- Depth of penetration = 0.038" Diablo Canyon (PWR) 9/16/86 Bechtel (1)
- RC slab, 24" thick 8.02 ft/sec
- Assumes slab thickness of 24"
- Assumes missile impact
- Depth of penetration = 0.038" Dresden 1 (BWR) 9/28/93 Bechtel (1)
TN-RAM cask 38.5 2'
above pool
- water, 41' of water
- Bottom of spent fuel pool
- RC2-3 ft. thick with rock base
- 44.9 ft/sec
- Bechtel Design Guide C-2.45, Design of Structures for Tornado Missile Impact, Rev 0"
- Drops vertically, lands totally flat on cask base
- The concrete base will fail in shear Dresden 1 (BWR) 9/28/93 Bechtel (2)
TN-RAM cask 38.5 2'
above pool
- water, 41' of water
- Bottom of spent fuel pool
- RC2-3 ft. thick with rock base
- 38.4 ft/sec
- Bechtel Design Guide C-2.45, Design of Structures for Tornado Missile Impact, Rev 0"
- Drops horizontally, contact area is calculated assuming a.76 inch penetration (1631 square inches)
- The concrete base will fail in shear Dresden 1 (BWR) 9/28/93 Bechtel (3)
Tn-9.1 Cask 41.5 2'
above pool
- water, 41' of water
- Bottom of spent fuel pool
- RC2-3 ft. thick with rock base
- 38.4 ft/sec
- Bechtel Design Guide C-2.45, Design of Structures for Tornado Missile Impact, Rev 0"
- Drops vertically, lands totally flat on cask base
- The concrete base will fail in shear
Table 8: Load drop calculations for very heavy components (continued)
PLANT CALC DATE LOAD WT (tons)
HT TARGET COMPOSITION STRIKING VELOCITY ASSUMPTIONS CALCULATION OUTCOMES 46 Dresden 1 (BWR) 9/28/93 Bechtel (4)
Tn-9.1 Cask 41.5 2'
above pool
- water, 41' of water
- Bottom of spent fuel pool
- RC2-3 ft. thick with rock base
- 29.0 ft/sec
- Bechtel Design Guide C-2.45, Design of Structures for Tornado Missile Impact, Rev 0"
- Drops horizontally, contact area is calculated assuming a.406 inch penetration (2044 square inches)
- The concrete base will fail in shear Dresden 1 (BWR) 10/6/93 Bechtel (5)
Tn-9.1 Cask 41.5 6"
(this dimens ion should have been at least 10 inches
- 8" thick RC wall
- 5.67 ft/sec
- Bechtel Design Guide C-2.45, Design of Structures for Tornado Missile Impact, Rev 0"
- The cask would have to go over the transfer pool curb which is 10 inches, not 6 inches as assumed in the calculation
- spalling will not occur since wall is >>31.9 inches
- Speculation is made for drops on the walkway next to the transfer pool Dresden 1 (BWR) 9/28/93 Bechtel (6)
TN-RAM cask 38.5 12"
- Washdown area floor
- 9" thick RC slab
- 8.02 ft/sec
- Bechtel Design Guide C-2.45, Design of Structures for Tornado Missile Impact, Rev 0"
- The concrete base will fail in shear Dresden 1 (BWR) 10/5/93 Bechtel (7)
TN-RAM 38.5
- See Dresde n (3) above
- See Dresden (3) above
- See Dresden (3) above
- See Dresden (3) above
- Assumes a redwood crush pad at the bottom of the spent fuel pool
- Assumes that the cask lands flat
- Acceptable (59% of allowable)
Dresden 1 (BWR) 10/5/93 Bechtel (8)
Tn-9.1 Cask 41.5
- See Dresde n (5) above
- See Dresden (5) above
- See Dresden (5) above
- See Dresden (5) above
- Assumes a redwood crush pad at the bottom of the spent fuel pool
- Assumes that the cask lands flat
- Acceptable (93% of allowable)
Dresden 1 (BWR) 10/6/94 Vectra (9)
Spent fuel casks75-110
- 3.75 ft. air, 39.25 water
- Fuel transfer slab, RC 3ft. thick
- Variable from approximat ely 38 to almost 47 ft/sec
- Bechtel Topical Report, Design of Structures for Missile Impact, BC-TOP-9A, Rev 2
- Modified Petry formula (missile penetration)
- Uses a Ballistic Research Lab formula
- Assumes a flat cask impact area (100%
contact) for all equations
- Punching shear is the controlling failure mode
- Acceptable for penetration, perforation and spalling (however, impact area of 100% was assumed)
- Spent fuel pool slab will fail by punching shear
- An energy absorbing device would have to be supplied to cover an area of 17 ft. x 10 ft.
(Even assuming a flat cask impact area)
Table 8: Load drop calculations for very heavy components (continued)
PLANT CALC DATE LOAD WT (tons)
HT TARGET COMPOSITION STRIKING VELOCITY ASSUMPTIONS CALCULATION OUTCOMES 47 Dresden 1 (BWR) 10/6/94 Vectra (10)
Spent fuel cask 110 NA
- Fuel transfer slab, RC 3ft. thick approximat ely 40-42 ft/sec
- Vertical drop
- A 45 degree crack will propagate from the outer edge of the cask and completely penetrate the pool floor
- Assumes a hole in the pool floor of approximately 154 square ft
- A coefficient of permeability (.0137 ft/day) for a sandy clay soil will be assumed
- Maximum leakage calculated to be approximately 2.7 gal/minute which should be easily made up by available water sources Dresden 2,3 (BWR) 5/21/73 S&L (11)
IF-300 GE cask 70 1.88 ft in air, 37.75 ft in water
- Spent fuel pool floor
- 6'-3" RC slab
- 44.1 ft/sec
- Vertical drop
- Modified Petry formula
- Assumes a flat cask impact (100% impact area of the fins, 445.5 square inches)
- Penetration in slab = 10.03"
- Load factor against punching shear = 2
- Lad factor against cracking = 1.44 Dresden 2,3 (BWR) 5/21/73 S&L (12)
IF-300 GE cask 70 1.88 ft in air, 37.75 ft in water
- Spent fuel pool floor
- 6'-3" RC slab
- 43.9 ft/sec
- Horizontal drop
- Modified Petry formula
- Assumes a reduce contact area of 1008 square inches
- Penetration in slab = 4.5"
- Load factor = 1.5
- Load factor against punching shear = 2 Dresden 2,3 (BWR) 5/21/73 S&L (13)
IF-300 GE cask 70 NA
- Decontamination pit NA
- Vertical drop
- Due to the complex shape, the slab was transformed into an equivalent fixed ended beam of 9.5' width
- The maximum drop height was calculated to be 11.15 inches
- It was recommended that the cask be raised a maximum of 9" for safe cleaning operation, and 6" while traveling to and from the decontamination pit Dresden 2,3 (BWR) 5/21/73 S&L (14)
IF-300 GE cask 70 NA
- Travel path between the decontamination pit and the spent fuel pool over the torus NA
- An extension form (13) above
- Vertical drop
- Two pathways were analyzed (slabs and beams, and over beams
- The pathway over beams was the most desirable, which indicated that the cask could be raised to a maximum height of 22."
- A conservative lift height of 6" was made Dresden 2,3 (BWR) 7/2/81 S&L (15)
NA NA 95.5 ft
- Drop down the reactor building equipment hatch to the main floor over the torus
- 78.4 ft/sec
- Assume the dropped load has a diameter of 18"
- RC slab, 24 inches thick
- Assumes concrete will fail at approximately 1300 Kips, then calculates the penetration depth into the concrete from an initial height of 95.5 ft
- To prevent scabbing of a 24" thick floor, the missile penetration depth cannot be > 3.27"
- Maximum load drop with no scabbing of a 24" thick slab = 1 ton
Table 8: Load drop calculations for very heavy components (continued)
PLANT CALC DATE LOAD WT (tons)
HT TARGET COMPOSITION STRIKING VELOCITY ASSUMPTIONS CALCULATION OUTCOMES 48 Dresden 2,3 (BWR) 7/2/81 S&L (16)
NA NA 95.5 ft
- Drop down the reactor building equipment hatch to the main floor over the torus
- 78.4 ft/sec
- Assume the dropped load has a diameter of 18"
- RC slab, 24 inches thick
- Assumes concrete will fail at approximately 1300 Kips, then calculates the penetration depth into the concrete from an initial height of 95.5 ft
- To prevent scabbing of a 24" thick floor, the missile penetration depth cannot be > 3.27"
- Maximum load drop with no perforation of a 24" thick slab = 5.75 tons Dresden 2,3 (BWR) 7/2/81 S&L (17)
NA NA 95.5 ft
- Drop down the reactor building equipment hatch to the main floor over the torus
- 78.4 ft/sec
- Assume the dropped load has a diameter of 18"
- RC slab, 32 inches thick
- Assumes concrete will fail at approximately 1300 Kips, then calculates the penetration depth into the concrete from an initial height of 95.5 ft
- To prevent scabbing of a 32" thick floor, the missile penetration depth cannot be > 3.27"
- Maximum load drop with no scabbing of a 32" thick slab = 2 tons Dresden 2,3 (BWR) 7/2/81 S&L (18)
NA NA
- Drop down the reactor building equipment hatch to the main floor, over the torus
- 78.4 ft/sec
- Assume the dropped load has a diameter of 24"
- RC slab, 32 inches thick
- Assumes concrete will fail at approximately 1300 Kips, then calculates the penetration depth into the concrete from an initial height of 95.5 ft
- To prevent scabbing of a 32" thick floor, the missile penetration depth cannot be > 4.62"
- The amount of energy deposit to produce a penetration depth of 4.62" was calculated to be 5.36 Kips
49 5
GUIDANCE FOR HEAVY LOAD MOVEMENTS AND CRANE CLASSIFICATION 5.1 Guidance from NUREG-0612 Section 5 of NUREG-0612, GUIDELINES FOR CONTROL OF HEAVY LOADS, states that:
Our evaluation of the information provided by licensees indicates that existing measures at operating plants to control the handling of heavy loads cover certain of the potential problem areas, but do not adequately cover the major causes of load handling accidents. These major causes included operator errors, rigging failures, lack of adequate inspection and inadequate procedures.
Section 5.1 of NUREG-0612, Recommended Guidelines, states that:
The objectives of these guidelines are to assure that either (1) the potential for a load drop is extremely small, or (2) for each area addressed, the following evaluation criteria are satisfied:
I.
Releases of radioactive material that may result from damage to spent fuel based on calculations involving accidental dropping of a postulated heavy load produce doses that are well within 10 CFR Part 100 limits of 300 rem thyroid, 25 rem whole body (analyses should show that doses are equal to or less than 1/4 of Part 100 limits);
II.
Damage to fuel and fuel storage racks based on calculations involving accidental dropping of a postulated heavy load does not result in a configuration of the fuel such that keff is larger than 0.95; III.
Damage to the reactor vessel or the spent fuel pool based on calculations of damage following accidental dropping of a postulated heavy load is limited so as not to result in water leakage that could uncover the fuel, (makeup water provided to overcome leakage should be from a borated source of adequate concentration if the water being lost is borated); and IV.
Damage to equipment in redundant or dual safe shutdown paths, based on calculations assuming the accidental dropping of a postulated heavy load, will be limited so as not to result in loss of required safe shutdown functions.
Section 5.1.1 of NUREG-0612, General, states that:
... all plants should satisfy each of the following for handling heavy loads that could be brought in proximity to or over safe shutdown equipment or irradiated fuel in the spent fuel pool area and in containment (PWRs), in the reactor building (BWRs), and in other plant areas.
50 (1)
Safe load paths should be defined for the movement of heavy loads...
(2)
Procedures should be developed to cover load handling operations for heavy loads...
(3)
Crane operators should be trained, qualified...
(4)
Special lifting devices should satisfy the guidelines...
(5)
Lifting devices that are not specially designed should be installed and used in accordance with the guidelines...
(6)
The crane should be inspected, tested, and maintained...
(7)
The crane should be designed to meet applicable criteria and guidelines...
To reverse the increase in the number of crane issues noted in Figures 1 and 8 in the last decade (1990 through 1999), and to prepare for an increase in the number of very heavy load lifts, it may be beneficial to revisit the guidelines presented above, and in NUREG-0612 in general. Many older nuclear facilities will reach the maximum capacity of their spent fuel pools, and are beginning to transfer, or will soon transfer, spent fuel to an independent spent fuel storage installation (ISFSI). Transfer of spent fuel to an ISFSI will increase the number of very heavy load lifts. In addition, as nuclear facilities are decommissioned, and spent fuel casks and other very heavy nuclear components are removed, a significant increase in the number of very heavy loads may also occur.
5.2 Licensee Response to NRC Bulletin 96-02 NRC Bulletin 96-02 was initiated in response to the planned movement of 100 ton dry storage casks by Oyster Creek during power operations. Based on the NRC audit of Oyster Creeks 10 CFR 50.59 evaluation of cask movement, the staff was concerned that other licensees may believe that their heavy load operations were in compliance with the regulations, because they had completed Phase I of the generic letter of December 22, 1980, and the closeout of Phase II by Generic Letter 85-11. In addition, Generic Letter 85-11 concluded that the risks associated with damage to safety-related equipment were relatively small because (1) nearly all load paths avoid this (safety-related) equipment, (2) most equipment is protected by an intervening floor, (3) there is redundancy or diversity of components, and (4) crane failure probability is generally independent of safety-related systems. As is demonstrated by Oyster Creeks proposed activities, and the information presented in the bulletin, this conclusion may not always be valid.
NRC Bulletin 96-02 requested licensees to provide the staff with specific information relating to their heavy loads program and plans within 30 days. Many of the licensees that responded to the bulletin, provided incomplete information relating to crane types, load drop analysis, consequence analysis, plant status during load movement, and crane type to be used for the load movements. Eight respondents indicated that a consequence analysis had been done at their facility for heavy load drops.
51 5.3 Single-failure-proof Crane Classification Single Failure Proof Crane Guidance NUREG-0554, Single-Failure-Proof Cranes for Nuclear Power Plants, and NUREG-0612, Control of Heavy Loads at Nuclear Power Plants, provide current NRC guidance for what constitutes design requirements for single-failure-proof cranes (NUREG-0554), or what modifications are required to upgrade an existing crane to a single-failure-proof classification (Appendix C of NUREG-0612). Both documents have been interpreted differently by licensees and vendors. It was also unclear what credit (waivers from standard or routine load handling requirements) could be given by the NRC to licensees that had modified cranes to make them more reliable and failure proof, when making very heavy load movements over safety-related equipment, or during power operations, if the crane did not meet all of the design criteria of NUREG-0554 or Appendix C of NUREG-0612.
ASME NOG-1, Rules for Construction of Overhead and Gantry Cranes (Top Running Bridge, Multiple Girder), received ANSI approval in October 1998. The NOG-1 Standard applies to the design, manufacture, testing, inspection, shipment, storage, and erection of cranes (Types I, II, and III) covered by the Standard. NOG-1, Type I crane design criteria appears to be similar to design criteria in NUREG-0554. The definition of a Type I crane in the NOG-1 Standard is:
a crane that is used to handle a critical load. It shall be designed and constructed so that it will remain in place and support the critical load during and after a seismic event, but does not have to be operational after this event.
Single failure-proof features shall be included so that any credible failure of a single component will not result in the loss of capability to stop and hold the critical load.
NOG-1 defines a critical load as, any lifted load whose uncontrolled movement or release could adversely affect any safety-related system when such a system is required for unit safety or could result in potential off-site exposure in excess of the limit determined by the purchaser.
Crane Classification Issues During the information gathering phase of this candidate GI, it became clear that definite criteria did not exist for declaring a crane as single-failure-proof (e.g., for new cranes or upgraded cranes). Crane manufacturers also stressed that NUREG-0554 was ambiguous in some areas, and that clarifications or changes needed to be made to both NUREG-0612 and NUREG-0554.
Industry suggested that a preferred approach would be to consider adopting NOG-1, Type I (with minor changes) as an acceptable approach to meeting NUREG-0554 and for upgrading cranes to single-failure-proof status. NOG-1 contains much more design information than NUREG-0554 in explaining design criteria for single-failure-proof cranes.
In addition, some licensees listed a crane as single-failure-proof, or that it met NUREG-0612 upgrade requirements, although all the design conditions in NUREG-0554 may not be fully met.
52 For example:
(1)
Oyster Creek Oyster Creek made many changes to their reactor building crane to increase its reliability, to reduce the likelihood of a load drop, and to minimize the consequence of a load drop to get approvals from the NRC to move dry storage casks at power, subsequently abandoned that approach, and installed a single-failure-proof crane in 2000.
(2)
Dresden The Dresden Unit 2/3 reactor building crane was accepted by the NRC as single-failure-proof (circa 1976) which was before NUREG-0554 was issued. Shortly after the licensing action of 1976, some of the features relied upon in classifying the crane as single-failure-proof were disabled, and generally remained so for many years. Restoration of the crane to its 1976 status, by installing new controls and limiting devices not bring the crane in full compliance with the current single-failure-proof standards of NUREG-0554.
Attachment E, Seismic Design of the Support Structures for the Reactor Building Crane, to Sargent and Lundy calculation DRE98-0020, Evaluation of Reactor Building Superstructure, dated March 16, 1998, provides a summary of reactor building design criteria. It states that calculations performed in 1966-1967 showed that the stresses in the girder, the support columns and several members of the roof truss were above the materials yield stress for the (dead weight plus safe shutdown earthquake) loading. In some of the roof truss connections, the (dead weight plus safe shutdown earthquake) loading exceeded the ultimate capacity of the connections. Calculations also do not include the stresses in the support column due to the seismic (operating basis earthquake or safe shutdown earthquake) loads imposed by the siding. In 1973, the 1966-67 calculations for the crane girder and the crane columns were updated (red marked) for the effects of the new heavier single failure proof trolley. The 1973 update shows that the columns are overstressed by 6% and 35% for the operating basis earthquake and safe shutdown earthquake loading respectively, with the heavier crane trolley.
In 1975 new calculations were prepared for the columns and the vertical bracing to compute the effects of the new trolley. Modifications for the columns and the vertical bracing were designed.
This calculation used seismic inputs and analysis methodology more conservative than that stated in the UFSAR. The modifications designed in this calculation were not implemented.
The Dresden calculation book index carries the notation Project canceled, calculation not approved.
Calculation DRE98-0020 (1) did not appear to include conservatisms for aging, (2) assumed that the as-built condition was the same as the design requirement, and (3) assumed that the compressive strength of affected concrete (fc ) was 4700 psi (this value is normally assumed to be 4000 or less).
Table 9, Dresden reactor building steel superstructure interaction summary, shows the Interaction Coefficient (IC) for selected critical members of the reactor building superstructure where the value for the IC was 0.90 or greater. This information is shown in more detail in Section 6, Summary and Conclusions. Since the IC = (actual stress)/(allowable stress), a value approaching or exceeding 1.0 may indicate an overstress situation.
53 Calculation DRE98-0020 provided bases for not including certain loads in the stress calculations. For example:
Snow Load (The snow load is not specifically mentioned in the UFSAR.) Section 3.8.4.1.2 of the UFSAR states that the design code which was used to govern the construction documents was the Uniform Building Code (UBC), 1964 edition. The 1964 UBC did not include live load in the formula for total lateral load. Based on this provision, snow load is not included in the seismic loading combinations (of this calculation).
Wind Load The UFSAR indicates that the wind velocity for all structures is 110 mph. Other wind loads (tornado, etc.) are not in the scope of the calculation. (Section 3.3.1.1.1 of the UFSAR indicates that the reactor building was designed to withstand winds of 170 mph, and Section 3.3.2.2.1 indicates that the reactor building is designed to withstand tornado winds to 300 mph and still safely shutdown.)
Operating Basis Earthquake and Safe Shutdown Earthquake Loads The original design basis included pertinent dead and live loads as well as the OBE (or SSE) seismic loads with the crane in any location, and without lifted load. (The calculation did consider a scenario of an SSE concurrent with a maximum lifted load, but considered this to be beyond design basis. The calculation also does not include wind loads with either an OBE or an SSE.)
54 Table 9: Dresden reactor building steel superstructure interaction summary Element description Dead loads Snow load Max lifted load Wind OBE SSE IC (1) Interior crane column member (W14x119/W24x145) (H/N/39-49)
Yes Yes Yes No No No 0.992 (1) Interior crane column member (W14x119/W24x145) (H/N/39-49)
Yes Yes Yes Yes No No 0.90 (2) Interior building column members (W24x145) (H/N/39-49)
Yes Yes Yes No No No 0.996 (2) Interior building column members (W24x145) (H/N/39-49)
Yes Yes Yes Yes No No 1.00 (3) Interior crane/building column base connections (H/N/39-49)
Yes No Yes No No Yes 0.95 (5) Exterior column base connections (rows 38 & 50, except rows H & N)
Yes No No No No Yes 0.97 (5) Exterior column base connections (rows 38 & 50, except rows H & N)
Yes No Yes No No Yes 0.96 (13) Roof truss members (double angles)
Yes No No No No Yes 0.90 (15) Roof truss members (plate girders)
Yes Yes Yes No No No 1.05 (15) Roof truss members (plate girders)
Yes No No No Yes No 0.95 (18) Roof truss connections (double angles)
Yes No No No No Yes 0.90 (23) Crane girder member Yes Yes Yes No No No 0.93 (23) Crane girder member Yes Yes Yes Yes No No 0.93 (24) Crane girder connections Yes Yes Yes Yes No No 0.98 6
CRANE OPERATING EXPERIENCE STUDIES Several crane studies have been performed to estimate failure probabilities, component reliability, root causes, and human factors issues. NUREG-0612 along with more recent studies are briefly discussed in Sections 6.1 though 6.6.
6.1 NUREG-0612, Control of Heavy Loads at Nuclear Power Plants NUREG-0612 was published by the Office of Nuclear Reactor Regulation (NRR) of the NRC in July 1980. This study was based on data available from (1) Occupational Safety and Health Administration (OSHA), involving root cause data on over 1000 crane accidents during an unspecified time period, (2) the Department of the Navy, involving 466 crane events occurring between February 1974 and October 1977, and (3) NRC Licensee Event Report involving 34 crane events occurring between July 1969 and July 1979. Multiple probabilities are given for
55 various scenarios, however, the study states, Based on the data collected from the Navy, it is expected that the probability of handling system failure for nuclear plant cranes will be on the order of between 10-5 and 1.5 x 10-4 per lift. This probability of failure was a best estimate since Navy crane data does not indicate how many lifts were actually performed, (i.e., only the number of problems has been quantified).
6.2 EEG-74, Probability of Failure of the TRUDOCK Crane System at the Waste Isolation Pilot Plant (WIPP)
EEG-74 was published by the Environmental Evaluation Group (EEG) of the New Mexico Institute of Mining and Technology in May 2000. The WIPP is located in southeast New Mexico in bedded salt at a depth of 650 meters. The repository is designed to contain 850,000 drum equivalents of contact-handled transuranic waste and 8000 canisters of remote handled transuranic waste. The contact handled waste will be shipped from various defense generator and storage sites in an NRC certified container called a TRUPACT II. The TRUDOCK system consists of two six ton cranes. Crane cable/hook breaks were initially based on relatively old (1970s) U.S. Navy data in NUREG-0612 which produced a failure rate of approximately 2.0x10-5 per demand. Further analysis resulted in an evaluation which produced a more realistic value of 2.5x10-6 per demand. The report also indicated that there was a 95%
likelihood that not more than one dropped load will occur in approximately 34 years. EEG-74 is included in Appendix B.
6.3 Department of Energy Study, Independent Oversight Special Study of Hoisting and Rigging Incidents Within the Department of Energy This study was performed by the Office of Oversight, Office of Environment, Safety and Health, U.S. Department of Energy (DOE) in October 1996. Equipment studied included cranes, forklifts, and other during the period from October 1, 1993 through March 31, 1996. The other category included manual and power-operated hoists, chainfalls, and block and tackle.
The report analyzed 66 relevant hoisting and rigging incidents occurred during the 30 month study period. Relevant was defined as: (1) an event occurring during hoisting and rigging operations, or the use of hoisting and rigging equipment, as defined in the U.S. Department of Energy Hoisting and Rigging Handbook, AND (2) one that resulted in unsafe or improper conditions that necessitated the immediate suspension of the hoisting and rigging operation for any period of time, led to a near miss, or caused an accident. Unfortunately, no listing of the relevant crane incidents was given, however, root causes of the crane incidents were listed, and are shown in Table 10, Root causes of crane incidents at DOE facilities. As seen by the table, most crane incidents at DOE facilities are related to human factors issues such as inattention to detail, work organization and planning, and programmatic areas rather than crane hardware failures or deficiencies. The DOE study is included in Appendix C.
56 Table 10: Root causes of crane incidents at DOE facilities Root Cause Percent Root Cause Percent Inattention to detail 20 Other human error 3
Work organization and Planning 18 Insufficient refresher training 3
Procedure not used or used incorrectly 9
Lack of procedure 2
Policy not adequately defined, disseminated, or enforced 9
Communication problem 2
Defective or inadequate procedure 9
Inadequate work environment 0
Inadequate administrative control 9
Inadequate supervision 0
Inadequate or defective design 5
Error in equipment or materials selection 0
Defective or failed part 5
Weather 0
Insufficient practice or hands-on experience 5
No training provided 0
Other management problem 3
6.4 California Department of Industrial Relations, Crane Accidents 1997 - 1999 The report was prepared by the Division of Occupational Safety and Health, California Department of Industrial Relations in May 2000. Data for the report was gathered from Federal OSHAs Office of Management Data Services (OMDS) Website. Data was also gathered from Micro-to-Host reports from the Integrated Management Information System (IMIS). The report states that from January 1, 1997 through December 31, 1999, the Division of Occupational Safety and Health learned of, or had reported to it, a total of 158 accidents involving a crane.
The report sorts the crane accidents by crane type, crane operator injuries, private sector vs.
public sector, construction vs. non-construction, and accident causation. No mention is made concerning crane failure rates. This report is included in Appendix D.
6.5 NUREG-1738, Technical Study of Spent Fuel Pool Accident Risk at Decommissioning Nuclear Power Plants NUREG-1738 was prepared by the Office of Nuclear Reactor Regulation and published in February 2001. This report states that for decommissioning plants:
For a non-single-failure-proof load handling system, the drop frequency of a heavy load drop is estimated, based on NUREG-0612 information, to have a mean value of 3.4x10-4 per year. The number of heavy load lifts was based on the NEI (Nuclear Energy Institute) estimate of 100 spent fuel shipping cask lifts per year, which probably is an overestimate. For plants with a single-failure-proof load handling system or plants conforming to the NUREG-0612 guidelines, the drop frequency is estimated to have a mean value of 9.6x10-6 per year, again for 100 heavy load lifts per year but using data from U.S. Navy crane experience.
57 Once the load is dropped, the analysis must then consider whether the drop significantly damages the SFP (spent fuel pool).
NUREG-1738, Appendix 2C, Structural Integrity of Spent Fuel Pool Structures Subject to Heavy Loads Drops, states that:
A loss-of-inventory from the SFP could occur as a result of a heavy load drop.
For single-failure-proof systems where load drop analyses have not been performed at decommissioning plants, the mean frequency of a loss-of-inventory caused by a cask drop was estimated to be 2.0x10-7 per year (assuming 100 lifts per year). For a non-single-failure-proof handling system where a load drop analysis has not been performed, the mean frequency of a loss-of-inventory event caused by a cask drop was estimated to be 2.1x10-5 per year. The staff believes that performance and implementation of a load drop analysis that has been reviewed and approved by the staff will substantially reduce the expected frequency of a loss-of-inventory event from a heavy load drop for either a single failure-proof or non-single-failure-proof system.
NUREG-1738, Appendix 2C is included in Appendix E to this report.
6.6 Navy Crane Events NUREG-1738 used Navy crane event data from December 1995 to May 1999 to modify NUREG-0612 equations to quantify the failure rate of lifting equipment. During the time period, there were 11 incidents which involved loads in excess of 20 tons. Four different accident types were recorded for the 11 events, (i.e., overload, damaged crane, load collision, and damaged load) most of which were caused by human factors (i.e., not following procedures or lack of skills). The Navy crane event data is included in Appendix F, Navy Crane Operating Experience.
58 7
REFERENCES 1.
U.S. Nuclear Regulatory Commission, memorandum, Proposed Generic Safety Issue -
Potential Risk and Consequences of Heavy Load Drops in Nuclear Power Plants, April 19, 1999.
2.
U.S. Nuclear Regulatory Commission, memorandum, Proposed Generic Safety Issue -
Potential Risk and Consequences of Heavy Load Drops in Nuclear Power Plants, May 27, 1999.
3.
U.S. Nuclear Regulatory Commission, NUREG-0554, Single-Failure-Proof Cranes for Nuclear Power Plants, May 1979.
4.
U.S. Nuclear Regulatory Commission, NUREG-0612, Control of Heavy Loads at Nuclear Power Plants, July 1980.
Appendix A Crane Events at U.S. Nuclear Power Plants 1968 through 1999
A-1 Introduction A review of crane documents in the NRCs Nuclear Document System (NUDOCS), events reported by individual licensees, through NRC documents and inspection reports, by vendors, and the public for the period 1968 through 1999 resulted in 297 different issues. Depending on the severity of each issue, each issue may be discussed in several documents. Most are administrative (not following a procedure, load path issues, noncompliance with technical specifications, inadequate crane operational testing prior to use, etc.) and few relate to problems encountered when lifting loads of approximately 30 tons or more. The data and resultant sorting is shown on Table A1, Reported crane issues at U.S. nuclear power plants.
Abbreviations for sorting categories are shown on Table A2, Sorting category abbreviations, and abbreviations for nuclear power plants are shown on Table A3, Plant name abbreviations.
Abbreviations used in Tables A1, A2, and A3 are located at the end of this appendix.
Sorting of Crane Issues To analyze crane issues, several general categories were established, most with several subcategories. Once this information was entered in the database, sorts were performed to look for trends and patterns.
Category 1: Plant and event report date Subcategories include; plant docket number (DOC), plant name (NAME), event report year (YR), event report month (MO), and whether the issue occurred when the plant had an operating license and after January 1980 (0612 OPER) (Post NUREG-0612).
Category 2: Crane type Subcategories include; reactor building (RB), polar (PC), auxiliary (AUX),
refueling/manipulator (MC), spent fuel pool (SFP), tower (TOW), mobile (MOB), and other (OTHR).
Category 3: Crane component deficiency Subcategories include; structure (STR), control (CONT), brakes (BRK), rails (RAIL),
fasteners (FAST), unknown or indeterminate (UNK), and none (NON).
Category 4: Reported administrative cause for event Subcategories include; not following procedures (NFP), poor procedures (PP), didnt test (NT), load path inadequacy (LPI), ventilation inadequacy (VT), maintenance (MT),
engineering (ENG), operations (OPS), unknown or indeterminate (UNK), and none (NON).
A-2 Category 5: Safety Implication of event Subcategories include; Death (DTH), injury (INJ), radiation release (RAD), load slip (LS), load drop (LD), crane component drop (above the hook) (CCD), equipment deficiency or damage (EQ), loss or partial loss of power (LPL), and none (NON).
Category 6: Load description for slip or drop events Subcategories include; Load description (component and weight) (LOADESC), and height of drop or slip (HEIGHT).
A-3 Table A1: Reported crane issues at U.S. nuclear power plants
A-4 Table A1: Reported crane issues at U.S. nuclear power plants (continued)
A-5 Table A1: Reported crane issues at U.S. nuclear power plants (continued)
A-6 Table A1: Reported crane issues at U.S. nuclear power plants (continued)
A-7 Table A1: Reported crane issues at U.S. nuclear power plants (continued)
A-8 Table A1: Reported crane issues at U.S. nuclear power plants (continued)
A-9 Table A1: Reported crane issues at U.S. nuclear power plants (continued)
A-10 Table A1: Reported crane issues at U.S. nuclear power plants (continued)
A-11 Table A1: Reported crane issues at U.S. nuclear power plants (continued)
A-12 Table A1: Reported crane issues at U.S. nuclear power plants (continued)
A-13 Table A1: Reported crane issues at U.S. nuclear power plants (continued)
A-14 Table A2: Sorting category abbreviations AUX Auxiliary crane BRK Brake deficiency CCD Crane component drop (above the hook)
COMP Crane component deficiency other than specific components listed CONT Electrical control part of crane DOC Plant Docket number DTH Death associated with event ENG Engineering personnel responsible EQ Equipment deficiency or damage FAST Fastener deficiency HEIGHT Approximate load slip or drop height INJ Injury associated with event LD Load drop (equipment damage or impact)
LOADDESC Description of load for slip or drop events LPI Load path inadequacy LPL Loss or partial loss of off-site power LS Load slip (equipment damage not incurred)
MC Manipulator crane MO Month of event record MOB Mobile crane MT Maintenance personnel responsible NAME Plant name NFP Not following procedure NON None or nothing effected NT No test or failed to test 0612 OPER Post NUREG-0612 operating facility OTHR Other crane (not specifically identified)
OPS Operations personnel responsible PC Polar crane PP Poor procedure RAD Radiation release RAIL Rail (for truck or trolley) deficiency RB Reactor building crane SFP Spent fuel pool crane STR Structural deficiency TOW Tower crane UNK Unknown or indeterminate VT Ventilation or ventilation test inadequacy Year Year of event record
A-15 Table A3: Plant name abbreviations ANO Arkansas Nuclear One BELL Bellefonte BF Browns Ferry BRA Braidwood BRU Brunswick BV Beaver Valley BYR Byron CAL Callaway CAT Catawba CC Calvert Cliffs CLI Clinton COO Cooper CP Comanche Peak CRY Crystal River DA Duane Arnold DB Davis-Besse DCC D.C. Cook DIC Diablo Canyon DRE Dresden FAR Joseph M. Farley FER Fermi FITZ James A. FitzPatrick FSV Fort St. Vrain FTC Fort Calhoun GG Grand Gulf GIN Ginna HART Hartsville HAT Edwin I. Hatch HN Haddam Neck IP2 Indian Point 2 IP3 Indian Point 3 LAS La Salle County MCG McGuire MH Marble Hill MIL Millstone MONT Monticello MY Maine Yankee NA North Anna NMP Nine Mile Point OC Oyster Creek OCO Oconee PAL Palisades PB Peach Bottom PER Perry PI Prairie Island PIL Pilgrim PTB Point Beach PV Palo Verde QC Quad Cities RB River Bend ROB H. B. Robinson RS Rancho Seco SAL Salem SEA Seabrook SEQ Sequoyah SH Shearon Harris SHO Shoreham SON San Onofre STP South Texas Project SUM Summer SUR Surry SUS Susquehanna TMI Three Mile Island TP Turkey Point TRO Trojan VOG Vogtle VY Vermont Yankee WC Wolf Creek WNP Washington Nuclear (Columbia)
ZIO Zion
A-16 1.
U.S. Nuclear Regulatory Commission, memorandum, Proposed Generic Safety Issue -
Potential Risk and Consequences of Heavy Load Drops in Nuclear Power Plants, April 19, 1999.
2.
U.S. Nuclear Regulatory Commission, memorandum, Proposed Generic Safety Issue -
Potential Risk and Consequences of Heavy Load Drops in Nuclear Power Plants, May 27, 1999.
3.
U.S. Nuclear Regulatory Commission, NUREG-0554, Single-Failure-Proof Cranes for Nuclear Power Plants, May 1979.
4.
U.S. Nuclear Regulatory Commission, NUREG-0612, Control of Heavy Loads at Nuclear Power Plants, July 1980.