Resolution of Generic Safety Issues: Issue 186: Potential Risk and Consequences of Heavy Load Drops in Nuclear Power Plants ( NUREG-0933, Main Report with Supplements 1–35 )
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DESCRIPTION
Historical Background
The Office of Nuclear Reactor Regulation (NRR) identified this issue in April 1999,1845 when the concern was raised that licensees operating within the regulatory guidelines of Generic Letter (GL) 85-11, “Completion of Phase II of ‘Control of Heavy Loads at Nuclear Power Plants,’” dated June 28, 1985," 1844 may not have taken adequate measures to assess and mitigate the consequences of dropped heavy loads. Before GL 85-11, the NRC issued GL 80 113, “Control of Heavy Loads,” on December 22, 1980,1842 GL 81-07, "Control of Heavy Loads," on February 3, 19811843 and GL 83-42, "Clarification to GL 81-07 Regarding Response to NUREG-0612, "Control of Heavy Loads at Nuclear Power Plants," on December 19, 1983.1846 These GLs gave requirements for operating licensees following the Resolution of Generic Safety Issues: Item A-36: "Control of Heavy Loads near Spent Fuel." In April 1996, the U.S. Nuclear Regulatory Commission (NRC) issued NRC Bulletin 96-02, “Movement of Heavy Loads over Spent Fuel, over Fuel in the Reactor Core, or over Safety-Related Equipment,”1848 to alert licensees of potential high risk and consequences that could result from a cask drop and to remind them to comply with existing regulatory guidelines on the control and handling of heavy loads.
Safety Significance
During nuclear plant operation, maintenance, and refueling activities, heavy loads may be handled in several plant areas, including areas with safety-related equipment. If these loads were to drop because of human error or crane failure, they could impact on stored spent fuel, fuel in the core, or on safety-related equipment that may be required to achieve safe shutdown or ensure decay heat removal. In some instances, load drops at specific times, locations, and weights could potentially lead to offsite doses that exceed the limits in Title 10 of the Code of Federal Regulation (10 CFR) Part 100, "Reactor site criteria".
Moreover, in 2003, as spent fuel pools were approaching their capacity, many licensees elected to use long-term dry storage casks to store excess spent fuel. In the process of loading fuel, the large, heavy casks would have to be hoisted and transported to and from the spent fuel pool. Typically, licensee perform this operation while the units are at full power operation.
In general, heavy load drops in boiling water reactors (BWRs) are more risk significant than heavy load drops in pressurized water reactors (PWRs) because of the layout of plant systems. For PWRs, spent fuel cask transfers occur near ground level in an area separate from the reactor building and safety-related systems required for safe shutdown of the operating units. However, for BWRs, many heavy loads are commonly lifted and moved on the upper floor of the reactor building or the auxiliary building. In the event a floor breach occurred during a load drop, there are many safety-related components located on lower floors that could be disabled. A load drop in certain critical areas could simultaneously initiate an accident and disable accident mitigation equipment. These types of events have the greatest potential to defeat defense-in-depth.
ANALYSIS
Frequency Estimate
The NRC conducted a comprehensive analysis of U.S. nuclear industry crane operating experience from 1968 through 2002 and documented the results in NUREG-1774, "A Survey of Crane Operating Experience at U.S. Nuclear Power Plants from 1968 through 2002," issued July 2003.1846 Some of the significant NRC's findings include the following:
(1) The human error rate for crane operating events increased significantly.
(2) Load drop events during 1993–2002 increased over the period 1981–1992.
(3) The number of below-the-hook crane events (mainly rigging deficiencies or failures) increased greatly.
(4) Calculational methodologies, assumptions, and predicted consequences varied greatly from licensee to licensee for very similar accident scenarios.
(5) The number of mobile crane events declined slightly.
(6) there were few load slips or drops involving very heavy loads.
Based on actual crane operating experience data from commercial U.S. nuclear power plants, the average rate of drops for very heavy loads was estimated at 5.6x10-5 per demand. This estimate could be higher or lower at a specific plant because of varying human error rates at that plant which appeared to dominate load drop events. Based on data estimates collected from the U.S. Navy, NUREG 0612, “Control of Heavy Loads at Nuclear Power Plants,” issued July 1980,747 estimated the frequency of a handling failure for nuclear plant cranes to be between 1x10-5 and 1.5x10-4 per lift. However, the Navy crane data did not indicate how many lifts were actually performed (i.e., only the number of problems was quantified).
Consequence Estimate
Of the 74 plants that responded to NRC Bulletin 96-02,1848 only eight indicated that a consequence analysis for heavy load drops had been done at their plants. While the number of operating power plants during 1993-2002 only increased by 9 percent over the previous period of 1981-1992, the number of crane-related injuries increased by 100 percent in that period. Between 1969 and 2002, 10 reported crane events that led to deaths in the nuclear industry. These deaths occurred primarily during the construction phase of the plants.
Other Considerations
The NRC documented the following observations in NUREG-1774
- Although single-failure-proof cranes share many common design features (e.g., dual reeving, redundant limit switches, and redundant brakes), the remaining criteria for declaring a crane as single-failure-proof (e.g., for new cranes or upgraded cranes) were applied inconsistently. Crane manufacturers were of the opinion that NUREG-0554 "Single-Failure-Proof Cranes for Nuclear Power Plants," issued May 1979,1849 was ambiguous in some areas and that clarifications or changes to both NUREG-0612 and NUREG-0554 were needed. The industry suggested that a preferred approach would be to consider adopting American Society of Mechanical Engineers (ASME) NOG-1,"Rules for Construction of Overhead and Gantry Cranes Type I," (https://www.asme.org/codes-standards/find-codes-standards/nog-1-rules-construction-overhead-gantry-cranes) with minor changes, as an acceptable approach to meeting the guidelines in NUREG-0554 and for upgrading cranes to single-failure-proof status. ASME NOG-1 contains much more specific design criteria for single-failure-proof cranes than does NUREG-0554. In addition, while some licensees listed their cranes as single-failure-proof or indicated that they met the NUREG-0612 upgrade requirements, all the single-failure-proof design criteria listed in NUREG-0554 may not have been fully met. Among events occurring during 1968-2002 involving cranes suitable for an upgrade to a single-failure-proof design, most load drop events were the result of poor program implementation or human performance errors that led to hoist wire rope or below-the-hook failures. All three very heavy load drops were the result of rigging failures, not crane failures. Consequently, there were no very heavy load drop events that could have been prevented had only a single-failure-proof crane been employed in the lift. However, there were load or hook and block assembly drops that could have been prevented with the use of single-failure-proof cranes and lifting devices.
- Between 1976 and 2003, the NRC issued 29 generic communications that involved heavy load movements at U.S. nuclear power plants, 9 of which addressed the following four issues:
- (1) heavy loads moved on the refueling floor
- (2) load drop analysis for heavy loads
- (3) the identification of heavy loads that are lifted over safe-shutdown equipment
- (4) the consequence of a load drop on selected equipment.
- These GLs and NRC Bulletin 96-02 requested licensees to provide information on their crane programs. However, the accuracy and consistency of information received in response to some of these communications were questionable. Many of the licensees provided incomplete information. In many instances, the staff found that information provided previously was not accurate.
RECOMMENDATIONS
- Evaluate the capability of various rigging components and materials to withstand rigging errors and issue necessary guidelines for rigging applications.
- Endorse ASME NOG-1 for Type I cranes as an acceptable method of qualifying new or upgraded cranes as single-failure-proof and issue guidance endorsing the standard, as appropriate.
- Reemphasize the need to follow Phase I guidelines involving good practices for crane operations and load movements and continue to assess licensee implementation of heavy load controls in safety-significant applications.
- Request the appropriate industry code committees to evaluate the need to standardize load drop calculational methodologies for nuclear power plants.
FOLLOW-UP ACTIONS TO THE RECOMMENDATIONS
Through its work with the ASME Cranes for Nuclear Facilities Committee, the NRC staff has concluded that the industry standard, ASME NOG-1, provides improved guidance for construction of new single-failure-proof cranes. Therefore, in NUREG-0800, “Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR Edition” (SRP), Section 9.1.5, “Overhead Heavy Load Handling Systems,” Revision 1, issued March 2007, the NRC staff endorsed the criteria for Type 1 cranes found in ASME NOG-1-2004 (https://infostore.saiglobal.com/en-us/standards/asme-nog-1-2004-138353_saig_asme_asme_296613/). The ASME Cranes for Nuclear Facilities Committee has since provided a comparison of the NRC version in the SRP to the original version in ASME NOG-1 as an appendix to the 2010 revision of ASME NOG-1. The NRC staff also modified the guidelines for slings used with single-failure-proof handling systems in SRP Section 9.1.5 based on a review of operating experience issues. The NRC staff issued Regulatory Issue Summary 2005-25, Supplement 1, “Clarification of NRC Guidelines for Control of Heavy Loads,” dated May 29, 20072218 to notify industry of the changes to SRP Section 9.1.5 and further clarify existing regulatory expectations associated with 10 CFR 50.59, “Changes, Tests and Experiments,” and 10 CFR 50.71(e), as these requirements relate to the safe handling of heavy loads and load drop analyses.1900 Based upon these actions recommendations (1) and (2) are resolved.
In a letter dated, September 14, 2007, "Subject: Industry Initiative on Heavy Load Lifts"2190, the Nuclear Energy Institute (NEI) notified the NRC that the nuclear industry approved a formal initiative that specifies actions each plant will take to ensure that heavy load lifts continue to be conducted safely and that plant licensing bases accurately reflect plant practices. The initiative is expected to clarify the licensing basis with respect to the handling of heavy loads, and the NRC staff is modifying guidance documents to accommodate the initiative. The initiative includes development of guidelines for realistic load drop analyses and for establishing single-failure-proof crane equivalence for reactor vessel head lifts.
On December 13, 2007,1992 February 1, 2008,1993 and April 8, 2008,2191 the NRC staff participated in public meetings with NEI to discuss implementation of the initiative and criteria for acceptable reactor vessel head load drop analyses.
On April 17, 2008,2192 the NRC staff participated in a public meeting with NEI to discuss draft guidelines for establishing single failure proof crane equivalence for reactor vessel head lifts. Therefore recommendation (3) is resolved.
By letters dated April 17, 2008,1994 and April 22, 2008 1995, NEI submitted the guidelines for reactor vessel head drop analyses and the guidelines for establishing single failure proof crane equivalence for reactor vessel head lifts, respectively. On May 16, 2008,1996 the NRC staff issued a letter providing preliminary endorsement of these guidelines with exceptions related to the load drop analysis acceptance criteria.
- Managing the risk associated with maintenance involving the movement of heavy loads
- Performing consequence analyses for postulated reactor vessel head drops
- Establishing single-failure-proof equivalence for handling systems when used for reactor vessel head lifts
- Updating the description of heavy load handling programs in the safety analysis report.
On September 5, 2008, the staff issued its safety evaluation of NEI 08-05,1999 endorsing the methods in NEI 08-05 for the specified applications, with certain exceptions and clarifications. On September 9, 2008, the NRC staff issued supplementary inspection guidance addressing implementation of the industry initiative as Operating Experience Smart Sample FY2007-03, “Crane and heavy lift inspection, supplemental guidance for IP-71111.2M,” Rev 2,2187which provided the following guidance.
1) Support NRC review of licensees’ activities related to heavy lifts and crane inspections.
2) Ensure safety by comparing current heavy lifting procedures with licensees’ commitments, including the industry-wide commitment to implement the initiative on control of heavy loads.
3) Document any inspection findings in the quarterly inspection report.
In addition, the NRC issued Regulatory Information Summary (RIS) 2008-28, "Endorsement of Nuclear Energy Institute Guidance for Reactor Vessel Head Heavy Load Lifts,", dated December 1, 20082000 to notify stakeholders of NRC endorsement of the guidelines in NEI 08-05. Based upon these actions, recommendation (4) is resolved.
The NRC staff issued a request for Advisory Committee on Reactor Safeguards (ACRS) review2001 of a draft closeout memorandum on September 7, 2011.2001 In a memorandum to the NRC Executive Director for Operations dated October 12, 2011,2002 the Executive Director for the ACRS reported that the ACRS considered the proposed closeout of GI-186 during the 587th meeting of the ACRS held October 6-8, 2011, and that the ACRS Committee had no objection to staff issuance of the proposed closeout.
CONCLUSION
In a memorandum transferring GI-186 to NRR for resolution (Ref. 3), the staff of the Office of Nuclear Regulatory Research made four recommendations. The NRR staff implemented these recommendations. Hence, the staff concludes that all actions have been implemented to provide reasonable assurance of future safe handling of heavy loads. In a memorandum to the NRC Executive Director for Operations dated January 27, 2012,2003 the Director of the NRR reported that GI-186 was closed.
Thus, this issue is RESOLVED.
REFERENCES
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