ML041600035
| ML041600035 | |
| Person / Time | |
|---|---|
| Site: | Palisades  |
| Issue date: | 06/16/2004 |
| From: | Stang J NRC/NRR/DLPM/LPD3 |
| To: | Domonique Malone Nuclear Management Co |
| stang J, NRR/DLPM, 415-1345 | |
| References | |
| TAC MC1862 | |
| Download: ML041600035 (12) | |
Text
June 16, 2004 Mr. Daniel J. Malone Site Vice President Palisades Nuclear Plant Nuclear Management Company, LLC 27780 Blue Star Memorial Highway Covert, MI 49043-9530
SUBJECT:
PALISADES PLANT - ISSUANCE OF AMENDMENT RE: PALISADES SPENT FUEL POOL CRANE UPGRADE (TAC NO. MC1862)
Dear Mr. Malone:
The U.S. Nuclear Regulatory Commission has issued the enclosed Amendment No. 215 to Facility Operating License No. DPR-20 for the Palisades Plant. The amendment approves an update of the Final Safety Analysis Report (FSAR) in response to your application dated January 29, 2004, as supplemented on May 14, and June 2, 2004.
The amendment grants approval to update the FSAR to reflect modifications to the fuel pool building crane (L-3 crane) main hoist upgrade to the new rated capacity of 110 tons and reflect the new single-failure-proof design. Specifically, the amendment approves the use of the L-3 crane as a single-failure-proof crane for below-the-hook loads up to 110 tons.
A copy of our related safety evaluation is also enclosed. The Notice of Issuance will be included in the Commissions biweekly Federal Register notice.
Sincerely,
/RA/
John F. Stang, Senior Project Manager, Section 1 Project Directorate III Division of Licensing Project Management Office of Nuclear Reactor Regulation Docket No. 50-255
Enclosures:
- 1. Amendment No. 215 to DPR-20
- 2. Safety Evaluation cc w/encls: See next page
Palisades Plant cc:
Robert A. Fenech, Senior Vice President Nuclear, Fossil, and Hydro Operations Consumers Energy Company 1945 Parnall Rd.
Jackson, MI 49201 Arunas T. Udrys, Esquire Consumers Energy Company 1 Energy Plaza Jackson, MI 49201 Regional Administrator, Region III U.S. Nuclear Regulatory Commission 801 Warrenville Road Lisle, IL 60532-4351 Supervisor Covert Township P. O. Box 35 Covert, MI 49043 Office of the Governor P. O. Box 30013 Lansing, MI 48909 U.S. Nuclear Regulatory Commission Resident Inspectors Office Palisades Plant 27782 Blue Star Memorial Highway Covert, MI 49043 Michigan Department of Environmental Quality Waste and Hazardous Materials Division Hazardous Waste and Radiological Protection Section Nuclear Facilities Unit Constitution Hall, Lower-Level North 525 West Allegan Street P.O. Box 30241 Lansing, MI 48909-7741 Michigan Department of Attorney General Special Litigation Division 525 West Ottawa St.
Sixth Floor, G. Mennen Williams Building Lansing, MI 48913 Manager, Regulatory Affairs Nuclear Management Company, LLC 27780 Blue Star Memorial Highway Covert, MI 49043 Director of Nuclear Assets Consumers Energy Company Palisades Nuclear Plant 27780 Blue Star Memorial Highway Covert, MI 49043 John Paul Cowan Executive Vice President & Chief Nuclear Officer Nuclear Management Company, LLC 700 First Street Hudson, WI 54016 Jonathan Rogoff, Esquire Vice President, Counsel & Secretary Nuclear Management Company, LLC 700 First Street Hudson, WI 54016 Douglas E. Cooper Senior Vice President - Group Operations Palisades Nuclear Plant Nuclear Management Company, LLC 27780 Blue Star Memorial Highway Covert, MI 49043 October 2003
- Date of SE NRR-058 OFFICE PDIII-1/PM PDIII-1/LA SPLB/SC EMEB/SC OGC PDIII-1/SC NAME JStang THarris JYerokun KManoly BPoole LRaghavan DATE 06/10 /04 06/09/04 06/10/04
- 06/03/04*
06/14/04 06/15/04
NUCLEAR MANAGEMENT COMPANY, LLC DOCKET NO. 50-255 PALISADES PLANT AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 215 License No. DPR-20 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Nuclear Management Company, LLC (the licensee), dated January 29, 2004, as supplemented on May 14, and June 2, 2004 complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commissions rules and regulations set forth in 10 CFR Chapter I; B.
The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.
There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public; and (ii) that such activities will be conducted in compliance with the Commissions regulations; D.
The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commissions regulations and all applicable requirements have been satisfied.
2.
Accordingly, by Amendment No. 215, Facility Operating License No. DPR-20 is hereby amended to authorize a change to the final safety analysis report (FSAR), as set forth in the license amendment application dated January 29, 2004, as supplemented on May 14, and June 2, 2004, and evaluated in the associated safety evaluation by the Commissions Office of Nuclear Reactor Regulation. The licensee shall update the FSAR to reflect the fuel pool building crane (L-3 crane) main hoist upgrade to the new rated capacity of 110 tons and reflect the new single-failure-proof design. Specifically, the amendment approves the use of the L-3 crane as a single-failure-proof crane for below-the-hook loads up to 110 tons by adding a description of this change, as authorized by this amendment, and in accordance with 10 CFR 50.71(e).
3.
This license amendment is effective as of the date of issuance and shall be implemented within 60 days.
FOR THE NUCLEAR REGULATORY COMMISSION
/RA/
L. Raghavan, Chief, Section 1 Project Directorate III Division of Licensing Project Management Office of Nuclear Reactor Regulation Date of Issuance:
SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 215 TO FACILITY OPERATING LICENSE NO. DPR-20 NUCLEAR MANAGEMENT COMPANY, LLC PALISADES PLANT DOCKET NO. 50-255
1.0 INTRODUCTION
By application to the U.S. Nuclear Regulatory Commission (NRC) dated January 29, 2004, as supplemented on May 14, and June 2, 2004, the Nuclear Management Company, LLC (NMC, the licensee), requested approval to update the Final Safety Analysis Report (FSAR) for the Palisades Nuclear Power Plant, to reflect the fuel pool building crane (L-3 crane) main hoist upgrade and single-failure-proof design.
The proposed amendment would approve the licensee to update the FSAR to reflect a change in the licensing basis for the handling of heavy loads using the L-3 crane main hoist.
Specifically, the proposed changes would credit the L-3 crane as a single-failure-proof design, meeting the guidelines of NUREG-0612, Control of Heavy Loads at Nuclear Power Plants and NUREG-0554, Single-Failure-Proof Cranes for Nuclear Power Plants, and the amendment would also approve use of the L-3 crane for below-the-hook loads up to 110 tons.
The supplemental letters dated May 14, and June 2, 2004, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the NRC staffs original proposed no significant hazards consideration determination as published in the Federal Register on March 1, 2004 (69 FR 9649).
2.0 REGULATORY EVALUATION
Title 10 of the Code of Federal Regulations, Part 50, Appendix A, General Design Criteria, General Design Criterion 2 (GDC-2), Design bases for protection against natural phenomena, requires that structures, systems, and components important to safety shall be designed to withstand the effects of natural phenomena without a loss of capability to perform their safety functions. GDC-4, Environmental and dynamic effects design bases, requires, in part, that structures, systems, and components important to safety shall be designed to accommodate the effects of missiles (i.e., a dropped load). GDC-61, Fuel storage and handling and radioactivity control, requires, in part, that fuel storage and handling systems are designed to prevent significant reduction in fuel storage coolant inventory under accident conditions (i.e.,
protection against draining the spent fuel pool following a dropped load). Section 9.1.5, Overhead Heavy Load Handling Systems, of NUREG-0800, NRC Standard Review Plan, references the guidelines of NUREG-0612, Control of Heavy Loads at Nuclear Power Plants, and NUREG-0554, Single-Failure-Proof Cranes for Nuclear Power Plants, for implementation of these criteria in the design of overhead heavy load handling systems.
The basis for the guideline in NUREG-0612 was to minimize the occurrence of the principal causes of load handling accidents and to provide an adequate level of defense-in-depth for handling of heavy loads near spent fuel and safe shutdown systems. Defense-in-depth is generally defined as a set of successive measures that reduce the probability of accidents and/or the consequences of such accidents. In the area of control of heavy loads, the emphasis is on measures that prevent load drops or other load handling accidents.
In NUREG-0612, the NRC staff provided regulatory guidelines for control of heavy load lifts to assure safe handling of heavy loads in areas where a load drop could impact on stored spent fuel, in the reactor core, or equipment that may be required to achieve safe shutdown or permit continued decay heat removal. In an unnumbered letter dated December 22, 1980, as supplemented by Generic Letter (GL) 81-07, Control of Heavy Loads, dated February 3, 1981, the NRC requested that all licensees describe the extent to which the guidelines of NUREG-0612 were satisfied at their facility and what additional modifications would be necessary to fully satisfy the guidelines. This request was divided into two phases (Phase I and Phase II) for implementation by licensees. Phase I guidelines address measures for reducing the likelihood of dropping heavy loads and provide criteria for establishing safe load paths; procedures for load handling operations; training of crane operators; design, testing, inspection, and maintenance of cranes and lifting devices; and analyses of the impact of heavy-load drops.
Phase II guidelines address alternatives to either further reduce the probability of a load-handling accident or mitigate the consequences of heavy-load drops. These alternatives include using a single-failure-proof crane for increased handling system reliability, employing electrical interlocks and mechanical stops for restricting crane travel to safe areas, or performing load drops and consequences analyses for assessing the impact of dropped loads on plant safety and operations. Criteria for design of single-failure-proof cranes were included in NUREG-0554. Appendix C to NUREG-0612 provided alternative criteria for upgrading the reliability of existing cranes to single-failure-proof standards.
NUREG-0554, Single-Failure-Proof Cranes for Nuclear Power Plants, dated May 1979, identifies features of design, fabrication, installation, inspection, testing, and operation of single-failure-proof overhead crane handling systems used for handling heavy loads. It recommends special single-failure-proof features in the crane hoisting machinery and reeving. Single-failure-proof means the cranes hoisting and breaking systems will safely retain the load following any single failure in them. The criteria provided in this NUREG will be used to evaluate whether the upgraded crane meets the criteria for classification as single-failure-proof.
In a letter dated August 26, 1983, the NRC staff approved Ederers Generic Licensing Topical Report EDR-1 (P) A, Ederer Nuclear Safety-Related eXtra-Safety And Monitoring (X-SAM)
Cranes, Revision 3, dated October 8, 1982, as an acceptable method for meeting the guidelines of NUREG-0554 and NUREG-0612. Appendices B and C of EDR-1 (P)-A identify the plant-specific information that is needed to verify a specific retrofitted cranes conformance with NUREG-0554 guidelines. Appendix B summarizes the plant-specific crane data supplied by Ederer. Appendix C summarizes the regulatory positions to be addressed by the applicant.
Licensees who incorporated the use of Ederers hoist and trolley into the design of a crane are to submit Appendices B and C to address how plant-specific application of the Ederer system satisfies the guidelines of NUREG-0612 and NUREG-0554.
GL 85-11, Completion of Phase II of Control of Heavy Loads at Nuclear Power Plants, NUREG-0612, dated June 28, 1985, dismissed the need for the NRC staff to review the Phase II responses received from licensees, based on the improvements observed during review of the phase I responses. However, GL 85-11 encouraged the licensees to implement actions they perceived to be appropriate to provide adequate safety.
In NRC Bulletin 96-02, Movement of Heavy Loads over Spent Fuel, Over Fuel in the Reactor Core, or Over Safety-Related Equipment, dated April 11, 1996, the NRC staff addressed specific instances of heavy load handling concerns and requested licensees to provide specific information detailing their extent of compliance with the guidelines and their licensing basis.
The existing licensing basis for Palisades includes various measures to defend against unacceptable consequences from potential equipment failures that result in a cask drop. The heavy-load paths and interlocks, which were used for the former L-3 crane, remain in place and are unchanged for the upgraded crane, and heavy loads will continue to be restricted from being moved over spent fuel in the fuel pool in accordance with plant procedures. While the cask-drop evaluation will no longer be required for the main hook which is being upgraded to single-failure-proof design, the auxiliary hook has not been proposed to be upgraded to a single-failure-proof design and thus, requires a load-drop analysis. The existing load-drop analysis (100 ton cask drop from main hook) is being retained as a bounding analysis for the auxiliary hook which has a rated capacity of 15 tons.
3.0 TECHNICAL EVALUATION
3.1 Background
The NRC staff has reviewed the licensees technical and regulatory analyses in support of its proposed license amendment. The licensee is proposing two changes: 1) to increase the capacity of the L-3 crane main hoist from 100 tons to 110 tons, and 2) to upgrade the crane to a single-failure-proof design that satisfies the guidelines of NUREG-0612 and NUREG-0554.
The fuel building crane is an indoor electric overhead traveling bridge, single-trolley crane, with a radio-controlled operator unit. Originally, the main hoist had a rated capacity of 100 tons.
The crane is now equipped with a main hoist rated at a 110-ton lifting capacity and an auxiliary hoist with a 15-ton lifting capacity. The licensees upgraded crane will be used in support of the loading of new dry fuel storage casks. Due to a change in the dry fuel storage system being used at the plant, the load that is to be carried by the main hoist has increased from 100 tons to 110 tons. In addition to the increase in cask weight, the dimension of the new cask has also changed.
The main hoist was initially not designed with single-failure-proof technology and, therefore, the plant had to be capable of withstanding the consequences of a postulated 100-ton transfer cask drop by installing an impact-limiting pad to protect the spent fuel pool structure. Since the new main hoist for the upgraded crane is of the single-failure-proof design, the cask drop analysis is no longer required for load drops from the main hoist. As a result the impact-limiting pads previously installed in the spent fuel pool to protect the pool structure from the postulated transfer-cask-drop accident during dry fuel storage operations is being eliminated. The current load-drop analysis, without the impact-limiting pads, bounds the load drop for the auxiliary hoist; therefore, the licensee has retained the analysis.
3.2 Single-Failure-Proof Load Handling System Upgrade The NRC staff reviewed the information provided by the licensee on the modified L-3 crane.
The information provided included: (1) a summary of the proposed change, background and technical analysis, (2) summaries of the engineering analysis performed in support of the crane upgrade, (3) tables prepared by the licensee comparing the L-3 crane main hoist and trolley designs with the requirements of NUREG-0554 and NUREG-612 and (4) Appendix B, Summary of Plant Specific Crane Data Supplied by Ederer Incorporated, and Appendix C, Summary of Regulatory Positions to be Addressed by the Applicant, of EDR-1.
The modified L-3 crane is a single-failure-proof crane designed by Ederer Incorporated in accordance with the NRC-approved Ederer Incorporated Report, EDR-1, Generic Licensing Topical Report. The upgraded crane utilizes the existing crane runway and box girder bridge assembly to support a new trolley. The upgrading of the crane involved replacing the trolley and associated motors, drives, and electrical controls, and evaluating the crane bridge and supporting structure for the additional load. The heavy-loads paths and interlocks, which were used with the former L-3 crane, remain in place and are unchanged.
The licensee provided a compliance table comparing L-3 crane main hoist and trolley design with the applicable regulatory requirements for single-failure-proof handling system. The regulatory requirements were from appropriate sections of NUREG-0612 and NUREG-0554.
Included is a discussion by the licensee on how the objectives and general guidelines of NUREG-0612 will be met with regards to: (1) establishment of safe load paths; (2) development of procedures; (3) training and qualification of crane operators; (4) selection of special lifting devices; (5) selection of slings; (6) inspection, testing and maintenance of cranes; and (7) application of standards to crane design. Key interlocks in combination with administrative control will continue to ensure that heavy loads will continue to be restricted from being moved over spent fuel in the fuel pool.
In accepting EDR-1 (P)-A for reference in plant-specific actions, the NRC staff noted that the acceptance applied only to the features described in the topical report, and did not constitute acceptance to the overhead-crane handling systems or the requirements which may be necessary to assure the safety application of the crane system within the nuclear power plant.
The plant-specific information required, as identified in Appendices B and C of EDR-1 (P)-A, was included in the licensee submittal dated January 29, 2004. The NRC staff reviewed the plant-specific information which was presented in table form, and compared the information to the guidelines of NUREG-055, and NUREG-0612, and the exceptions to those guidelines (e.g., wire rope breaking strength) approved in the safety evaluation of EDR-1 (P)-A.
The NRC staff reviewed the plant-specific information related to the design of the hoist, the adequacy of specific components, the response of the crane to potential component failures and the test information demonstrating satisfactory performance of the overall crane. The NRC staff found that the plant-specific information provided adequate assurance that performance of the crane would satisfy the objectives of NUREG-0612 and the intent of NUREG-0554 with regard to maintaining the potential for a load drop extremely small.
3.3 Load Drop Considerations The licensee states in its January 29, 2004, application that [T]he use of a single-failure-proof hoist, together with rigging and procedures that are in compliance with the appropriate regulatory requirements, assures that the potential for a heavy-load drop is extremely small and therefore, not credible. As a result, a cask drop is not required to be postulated. The existing analysis of a cask drop will continue to serve as the bounding analysis for a postulated load drop from the auxiliary hoist, which is not single-failure proof. While it is clear that the existing drop analyses are bounding in terms of radiological consequences for a load drop, it is not clear whether or not the impact-limiting pads would be needed to protect the pool liner in the case of a drop from the auxiliary hoist. The NRC staff requested that the licensee provide additional information on the design and licensing basis for the 15-ton auxiliary hook. In its response by letter dated May 12, 2004, the licensee stated that a 25-ton load drop analysis had been performed for specific locations in the spent fuel pool. The analyses demonstrated that the spent fuel integrity is maintained for the worst-case load drop from the 15-ton auxiliary hoist without load-limiting pads. Therefore, based on the analysis the impact-limiting pads are not required for drops from the spent fuel pool crane auxiliary hoist. Since the probability of a load drop from the upgraded main hoist is extremely small, and the licensees 25-ton load drop analysis showed that the pool liner will not be damaged by a load drop from the auxiliary hook, the NRC staff finds the removal of the impact-limiting pads to be acceptable.
3.4 Structural Analysis The NRC staffs review of the crane structural analysis is based on Standard Review Plan (SRP, or NUREG-0800) Section 9.1.5, Overhead Heavy Load Handling Systems. This SRP section refers to NUREG-0612, Control of Heavy Loads at Nuclear Power Plants, and NUREG-0544, Single-Failure-Proof Cranes for Nuclear Power Plants. In accordance with the guidance in NUREG-0612, if a crane meets the criteria in NUREG-0544, a load drop is not considered to be a credible event.
The NRC staffs review focused on the licensees structural analyses of the load-bearing crane members and the auxiliary building support structures. The licensees January 29, 2004, submittal lists the series of calculations that support the proposed amendment. In response to the NRC staffs request, by letter dated May 14, 2004, the licensee submitted a summary of the structural analyses.
The crane was designed and evaluated in accordance with Crane Manufacturers Associates of America (CMAA) Specification 70, Specifications for Top Running Bridge and Gantry Type Multiple Girder Electric Overhead Traveling Crane. The application states that the design considered the load combinations in American Society of Mechanical Engineers (ASME)
NOG-1, Rules for Construction of Overhead and Gantry Cranes (Top Running Bridge, Multiple Girder), Section NOG-4140, Load Combinations.
The analysis used three-dimensional models of the structural steel framing and crane to generate the seismic response spectra of the bridge girder with the trolley and lifted loads in various locations and elevations. FSAR seismic accelerations were used. Pendulum loads were considered, with the pendulum frequency depending on the elevation of the lifted load.
The allowable stress limits for the crane are taken from CMAA, Specification 70, reflecting the appropriate duty cycle, which is "Service Level A - Standby or Infrequent Service" as defined by CMAA, Specification 70. Since CMAA, Specification 70, does not include seismic load combinations or allowable values, the NOG-I load combinations were used and the stresses were evaluated to the American Institute of Steel Construction design methodologv and stress limits.
The NRC staff requested additional clarification of the load cases for the structural steel (above the level of the spent fuel pool), the stress increase factors, and changes to the licensing bases methodologies. The licensee provided its response by letter dated June 2, 2004. The licensee stated that the load cases were derived from the Palisades FSAR, Section 5.9.1.1.2, Loads and Load Combinations, with the additional loads caused by a loaded crane, which is consistent with ASME NOG-1. Stress increase factors were used to normalize the equations based on allowable stress. The analyses demonstrated that one steel structural member would be overstressed during a seismic event. Therefore, the applicant modified the overstressed member such that the resulting stresses would be acceptable. The analyses also demonstrated that the auxiliary building structure would be overstressed if the ground-level wind speed exceeds 90 mph (Palisades FSAR design criteria is 100 mph) while a load is being lifted.
Therefore, the licensee incorporated an administrative control to specify that the L-3 crane will not be used when ground-level wind speeds are expected to exceed 90 mph (e.g., a weather warning is in effect).
3.5 Summary On the basis of preceding discussions, the NRC staff finds that the proposed upgrade of the existing spent fuel pool crane to a single-failure-proof design is in accordance with NUREG-0612 and satisfies the intent of NUREG-0554. The staff finds that the use of the proposed crane, with special lifting devices and slings meeting the specified design criteria, will enable the licensee to handle heavy loads with little risk to irradiated fuel stored in the spent fuel pool or redundant trains of safe shutdown equipment. The NRC staff finds that the licensee has used appropriate assumptions, loads, load combinations, seismic input spectra, and analytical methodology for performing the structural analyses to support increasing the L-3 crane capacity to 110 tons and qualifying the L-3 crane to single-failure-proof criteria.
Therefore, the NRC finds the changes to the FSAR concerning the modifications of the spent fuel pool crane are acceptable and pursuant to GDC 2, 4, and 61.
4.0 PUBLIC COMMENT By letters dated April 29 and March 31, 2004, the Office of the Secretary of the Commission received comments from Utilities Service Alliance, Inc., and the Strategic Teaming and Resource Sharing, respectively, on the Notice of Consideration of Issuance of Amendment to Facility Operating License and Opportunity for a Hearing, as published in the Federal Register on March 1, 2004 (69 FR 9649). Neither letter requested a hearing nor a to petition for leave to intervene. The comments focused on the licensees decision to seek approval from the NRC with a license amendment in lieu of performing an evaluation pursuant to 10 CFR 50.59, Changes, tests, and experiments. Because the comments did not specifically pertain to the licensing action at issue, the NRC will not further discuss the public comments in this safety evaluation.
5.0 STATE CONSULTATION
In accordance with the Commissions regulations, the Michigan state official was notified of the proposed issuance of the amendment. The Michigan state official had no comments.
6.0 ENVIRONMENTAL CONSIDERATION
Pursuant to 10 CFR 51.21, 10 CFR 51.32, and 10 CFR 51.35, an environmental assessment and finding of no significant impact was published in the FR on June 16, 2004 (69FR33686), for this amendment. Accordingly, based upon the environmental assessment, the Commission has determined that issuance of this amendment will not have a significant effect on the quality of the human environment.
7.0 CONCLUSION
The Commission has concluded, based on the considerations discussed above, that: (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commissions regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.
Principal Contributor: A. Stubbs S. Bailey Date: June 16, 2004