RS-03-171, Request for License Amendment Related to Heavy Loads Handling

Request for License Amendment Related to Heavy Loads Handling

ML032550306
Person / Time
Site:	Dresden
Issue date:	08/29/2003
From:	Simpson P Exelon Generation Co, Exelon Nuclear
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
RS-03-171
Download: ML032550306 (14)

Text

Exe n SM Exelon Generation www.exeloncorp.com Nu e 4300 Winfield Road Warrenville, IL 60555 10 CFR 50.59 10 CFR 50.90 RS-03-171 August 29, 2003 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Dresden Nuclear Power Station, Units 2 and 3 Facility Operating License Nos. DPR-19 and DPR-25 NRC Docket Nos. 50-237 and 50-249

Subject:

Request for License Amendment Related to Heavy Loads Handling

References:

(1) Letter from Keith R. Jury (Exelon Generation Company, LLC) to U. S.

NRC, "Request for License Amendment Related to Heavy Loads Handling," dated September 26, 2002 (2) Letter from U. S. NRC to J. L. Skolds (Exelon Generation Company, LLC), "Dresden - Issuance of Exigent Amendments - Lifting Heavy Loads," dated October 4, 2002 (3) Letter from P. R. Simpson (Exelon Generation Company, LLC) to U. S.

NRC, "Request for License Amendment Related to Heavy Loads Handling," dated February 26, 2003 In accordance with 10 CFR 50.90, "Application for amendment of license or construction permit,' and 10 CFR 50.59, "Changes, tests, and experiments," Exelon Generation Company, LLC (EGC) is requesting changes to Facility Operating License Nos. DPR-1 9 and DPR-25, for Dresden Nuclear Power Station (DNPS), Units 2 and 3. The proposed changes will allow EGC to use the Unit 2/3 reactor building crane during power operations to lift heavy loads in excess of 110 tons. Specifically, DNPS is requesting approval to revise the DNPS Updated Final Safety Analysis Report (UFSAR) to use the reactor building crane for heavy loads up to a total of 117 tons for removal and re-installation activities for the reactor shield blocks prior to and during the Unit 2 refueling outage D2R18. Reactor shield block removal activities are scheduled to commence on October 13, 2003.

In Reference 1, EGC requested a similar license amendment on an exigent basis to support refueling operations for DNPS, Unit 3. This request was intended to be applicable only for the October 2002 Unit 3 refueling outage and was necessary due to circumstances described in Reference 1. The NRC approved this request in Reference 2. The basis for the enclosed request is essentially the same as the basis for the Reference 1 amendment request.

In Reference 3, EGC submitted a license amendment request to revise the DNPS UFSAR to include a description of a load drop analysis performed for handling reactor cavity shield

August 29, 2003 U. S. Nuclear Regulatory Commission Page 2 blocks weighing greater than 110 tons with the Unit 2/3 reactor building crane during power operation. EGC requested that this amendment request be approved in time to support refueling outage D2R18, scheduled for October 2003. The NRC is currently reviewing this amendment request.

Following teleconferences with the NRC during the week of August 25, 2003, EGC determined that there was a potential that Reference 3 would not -be approved in time to support D2R1 8. Since removal of the reactor cavity shield blocks is necessary to accomplish refueling, EGC is requesting that the NRC approve the enclosed license amendment request if the Reference 3 amendment request cannot be approved by October 10, 2003. EGC intends to formally withdraw the enclosed amendment request if the Reference 3 amendment request is approved by October 10, 2003.

This request is subdivided as follows.

1. provides an evaluation supporting the proposed changes.

2. provides the proposed revisions to the UFSAR.

These proposed changes have been reviewed by the DNPS Plant Operations Review Committee and approved by the Nuclear Safety Review Board in accordance with the requirements of the EGC Quality Assurance Program.

EGC is notifying the State of Illinois of this request by transmitting a copy of this letter and its attachments to the designated State Official.

Should you have any queitions concerning his letter, please contact Mr. Allan R. Haeger at (630) 657-2807.

I declare under penalty of perjury that the foregoing is true and correct. Executed on the 29th day of August 2003.

Respectfully, Patrick R. Simpson Manager, Licensing : Evaluation of Proposed Changes : Proposed Revisions to the UFSAR cc:

Regional Administrator - NRC Region IlIl NRC Senior Resident Inspector - Dresden Nuclear Power Station Office of Nuclear Facility Safety - Illinois Department of Nuclear Safety

Attachment I EVALUATION OF PROPOSED CHANGES

1.0 INTRODUCTION

In accordance with 10 CFR 50.90, "Application for amendment of license or construction permit," and 10 CFR 50.59, "Changes, tests, and experiments," Exelon Generation Company, LLC (EGC) is requesting changes to Facility Operating License Nos. DPR-19 and DPR-25, for Dresden Nuclear Power Station (DNPS), Units 2 and 3. The proposed changes will allow EGC to use the Unit 2/3 reactor building crane during power operations to lift heavy loads in excess of 110 tons. Specifically, DNPS is requesting approval to revise the DNPS Updated Final Safety Analysis Report (UFSAR) to allow use of the crane for heavy loads up to a total of 117 tons for removal and installation activities for the reactor shield blocks prior to and during the Unit 2 refueling outage D2R18. Reactor shield block removal activities are scheduled to commence on October 13, 2003.

2.0 DESCRIPTION

OF PROPOSED AMENDMENT EGC is proposing to revise the DNPS UFSAR to allow use of the reactor building crane for lifting loads of up to 117 tons to support D2R18. A marked-up copy of the UFSAR has been provided as Attachment 2, detailing these changes. The total lifting time of these reactor shields blocks for both removal and reinstallation activities is estimated to be less than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

3.0 BACKGROUND

In Reference 1, EGC requested a similar license amendment on an exigent basis to support refueling operations for DNPS, Unit 3. This request was intended to be applicable only for the October 2002 Unit 3 refueling outage and was necessary due to circumstances described in Reference 1. The NRC approved this request in Reference 2.

In Reference 3, EGC submitted a license amendment request to revise the DNPS UFSAR to include a description of a load drop analysis performed for handling reactor cavity shield blocks weighing greater than 110 tons with the Unit 2/3 reactor building crane during power operation. EGC requested that this amendment request be approved in time to support DNPS Unit 2 refueling outage D2R18, which is scheduled for October 2003. The NRC is currently reviewing this amendment request.

Following teleconferences with the NRC during the week of August 25, 2003, EGC determined that there was a potential that Reference 3 would not be approved in time to support D2R1 8.

Since removal of the reactor cavity shield blocks is necessary to accomplish refueling, EGC is requesting that the NRC approve the enclosed license amendment request if the Reference 3 amendment request cannot be approved by October 10, 2003. -EGC intends to formally withdraw the enclosed amendment request if the Reference 3 amendment request is approved by October 10, 2003.

DNPS uses the reactor building crane for heavy loads to support refueling activities. The DNPS common refuel floor was originally designed to completely disassemble both Unit 2 and Unit 3 reactors simultaneously with all equipment stored within the boundaries of each unit. While this is an option for an emergency shutdown, eventual decommissioning or safe store operations, it is impractical to limit the laydown space to the shutdown unit for general refueling operations. Sharing of common equipment, such as the refuel bridges, Page 1 of 7

Attachment I EVALUATION OF PROPOSED CHANGES decontamination pad and the equipment hatch is required. The amount of equipment and resources that will be needed during the refueling outage will require all available floor space.

All laydown areas have been carefully orchestrated to allow free movement of refuel and specialty tooling as to not impede outage critical path activities and minimize crane moves because of large equipment obstructions. Utilizing all available refuel floor space optimizes time, which translates to increased personnel safety due to less restrictive work areas and lower dose rates due to better as low as reasonably achievable (ALARA) practices.

EGC is requesting this license amendment to allow DNPS to perform required activities as described above for its planned refueling outage. Since the reactor shield blocks are placed on the refuel floor of the operating unit (i.e., Unit 3), the requested amendment is needed to prevent a shutdown of Unit 3 to support D2R18. In addition, the requested amendment is needed to allow removal of Unit 2 reactor shield blocks during power operations.

The actual weight, based on measurement, of the top layer of shield blocks for DNPS, Unit 3 is less than 116 tons, including the rigging-used for lifting. The weight of the DNPS, Unit 2 reactor shield blocks is not known precisely. Based on a review of the dimensional drawings, DNPS has determined that the weight of the Unit 2 reactor shield blocks, including rigging, is expected to be less than 116 tons, and is unlikely to exceed 117 tons. DNPS will measure the weight of the Unit 2 reactor shield blocks and verify the weight is less than or equal to 117 tons before commencing movement of the reactor shield blocks. The second and third layers of reactor shield blocks are smaller and weigh less than the top layer.

4.0 REGULATORY REQUIREMENTS & GUIDANCE Regulatory guidance provided 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 1996, provides that movement of heavy loads over spent fuel, fuel in the reactor core, or safety related equipment while the reactor is at power should be conducted in accordance with applicable regulatory requirements and within the guidelines of the current licensing basis.

In NUREG-0612, "Control of Heavy Loads at Nuclear Power Plants," dated July 1980, the NRC provided regulatory guidelines in two phases (Phase I and 11) to assure safe handling of heavy loads in areas where a load drop could impact stored spent fuel, fuel in the reactor core, or equipment that may be required to achieve safe shutdown or permit continued decay heat removal. 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 for mitigating the consequences of heavy load drops, including using either (1) a single failure-proof crane for increased handling system reliability, or (2) electrical interlocks and mechanical stops for restricting crane travel, or (3) load drops and consequence analyses for assessing the impact of dropped loads on plant safety and operations. NUREG-0612, Appendix C provides alternative means of upgrading the reliability of the crane to satisfy the guidelines of NUREG-0554, "Single-Failure-Proof Cranes for Nuclear Power Plants."

Generic Letter (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 licensees to Page 2 of 7

Attachment I EVALUATION OF PROPOSED CHANGES implement the guidelines of NUREG-0612 Phase II based on the improvements obtained from the implementation of NUREG-0612 Phase I. GL 85-11, however, encouraged licensees to implement actions they perceived to be appropriate to provide adequate safety.

5.0 TECHNICAL ANALYSIS

EGC has concluded that the requested amendment is acceptable for the following reasons.

The reactor building crane was modified with the intent of qualifying it as single failure-proof for 125 tons. The reactor building crane has additional capacity for a total lifted load of 117 tons with single failure-proof features if a Design Basis Earthquake (DBE) is not assumed.

The probability of a DBE during the limited duration of the request is very small.

Reactor building crane capacity The stresses experienced by the DNPS reactor building crane were analyzed for the bridge, the trolley, and all of the major components listed in Attachment I of Reference 4. The various components have been designed with significant margin to the yield or ultimate strength of the material.

However, the licensing basis for this crane limits its load to 110 tons as a single failure-proof crane. If the DBE loads applied to the crane structures are removed from those structures (i.e., the bridge girders and the trolley), this results in a minimum increase of 10% in the load carrying capacity of these crane structures using the same allowables. This additional increase is more than enough to offset the lifted load increase of the crane to 117 tons.

A review of References 4 and 5 identifies that the factors of safety for the 125 ton reactor building crane single element components within the crane hoisting system load path and components critical to crane operations will increase by approximately 6% when the crane load is restricted to lifting 117 tons. Hence additional margin in the load carrying capacity of critical components will result.

On January 29, 2003, Whiting Corporation, the crane manufacturer, issued a report in accordance with 10 CFR 21, "Reporting of Defects and Noncompliance." This report identified design discrepancies regarding the load carrying capacity of certain components in some overhead cranes supplied by Whiting. The report recommended that crane owners review their crane designs and upgrade these components if necessary. DNPS has completed the review of the Whiting report and has replaced all components identified in the review, such that the DNPS reactor building crane has been restored to full load-carrying capacity.

The other features of the crane recognized by the NRC in approving the DNPS reactor building crane as single failure-proof are unaffected by this request. The crane hoist system consists of a dual load path through the hoist gear train, the reeving system, and the hoist load block along with restraints at critical points to provide load retention and minimization of uncontrolled motions of the load in the event of failure of any single hoist component.

Redundancy has been designed into the hoist and trolley brakes and the crane control components.

Page 3 of 7 EVALUATION OF PROPOSED CHANGES Probability of a Design Basis Earthquake Based on seismic estimates for the DNPS site that the NRC has published in NUREG-1488, "Revised Livermore Seismic Hazard Estimates for Sixty-Nine Nuclear Power Plant Sites East of the Rocky Mountains, 1994," the frequency of equaling or exceeding the DNPS DBE level is very low. Furthermore, as discussed above, the cumulative period of time required for the load lifts of concern is short (i.e., 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />). Therefore, the probability is very low that a DBE would occur during one of the load lifts.

6.0 REGULATORY ANALYSIS

The DNPS reactor building crane has been approved by the NRC as meeting single failure-proof criteria for handling heavy loads of up to 110 tons. The current DNPS UFSAR does not consider any credible load drop accidents that result from handling reactor shield plugs with the DNPS Unit 213 reactor building crane over safety-related equipment while the reactor is at power. Thus, since the crane is only approved as single failure-proof for loads of up to 110 tons, the proposed use of the crane for the activities described above could have the potential to create a new accident not analyzed in the UFSAR. This would require NRC approval in accordance with 10 CFR 50.59, "Changes, tests, and experiments." However, as stated in Section 7.0, we have concluded that the proposed changes involve no significant hazards consideration.

7.0 NO SIGNIFICANT HAZARDS CONSIDERATION According to 10 CFR 50.92, "Issuance of amendment," paragraph (c) a proposed amendment to an operating license involves no significant hazards consideration if operation of the facility in accordance with the proposed amendment would not:

(1)

Involve a significant increase in the probability or consequences of an accident previously evaluated; or (2)

Create the possibility of a new or different kind of accident from any accident previously evaluated; or (3)

Involve a significant reduction in a margin of safety.

In support of this determination, an evaluation of each of the three criteria set forth in 10 CFR 50.92 is provided below regarding the proposed license amendment.

Overview In accordance with 10 CFR 50.90, "Application for amendment of license or construction permit,"

Exelon Generation Company, LLC (EGC)is requesting changes to Facility Operating License Nos. DPR-19 and DPR-25, for Dresden Nuclear Power Station (DNPS), Units 2 and 3.

Specifically, the proposed changes will allow EGC to revise the DNPS Updated Final Safety Analysis Report (UFSAR) to allow use of the reactor building crane at DNPS during power operations to lift heavy loads up to a total of 117 tons for removal and re-installation activities for the reactor shield blocks prior to and during the Unit 2 refueling outage.

Page 4 of 7

Attachment I EVALUATION OF PROPOSED CHANGES The proposed changes do not Involve a significant increase In the probability or consequences of an accident previously evaluated.

The current DNPS licensing basis does not consider a load drop accident involving the reactor building crane as a credible event for loads up to and including 110 tons. The proposed changes will allow use of the reactor building crane at DNPS during power operations to lift heavy loads up to 117 tons for removal and installation activities for the reactor shield blocks prior to and during the Unit 2 refueling outage (i.e.' D2R 18). The reactor building crane has additional margin for a total lifted load of 117 tons with single failure-proof features if a Design Basis Earthquake (DBE) is not assumed. EGC has qualitatively demonstrated that the probability of a DBE occurring during the limited 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> duration of the request is very small.

The probability of load drop accidents is not increased since the single-failure proof capacity of the reactor building crane exceeds the weight of the reactor shield blocks, assuming that no DBE occurs. Since no load drop is assumed to occur, the consequences of a load drop accident are not affected.

Therefore, the proposed changes do not involve a significant increase in the probability or consequences of an accident previously evaluated.

The proposed changes do not create the possibility of a new or different kind of accident from any accident previously evaluated.

The proposed changes allow use of the DNPS reactor building crane for a limited duration to lift heavy loads up to a total of 117 tons during removal and installation activities for the reactor shield blocks. The reactor building crane has additional margin for a lifted load of 117 tons with single failure-proof features if a DBE is not assumed. The probability of a DBE during the limited duration of the request is very small. Therefore, the single failure-proof features ensure that the proposed changes provide an equivalent level of safety and will not create the possibility of a new or different kind of accident from any accident previously evaluated.

The proposed changes do not Involve a significant reduction in a margin of safety.

The reactor building crane is rated for lifting loads up to 125 tons. The NRC has approved qualification of the DNPS reactor building crane as single failure-proof for loads of up to 110 tons. The proposed change allows use of the crane for a limited duration to lift loads up to 117 tons. Existing safety margins are enhanced when lifting loads up to 117 tons if a DBE is not assumed, and EGC has demonstrated that the probability of a DBE during the limited duration of the request is very small. Therefore, it is concluded that the proposed changes do not result in a significant reduction in the margin of safety.

Conclusion Based upon the above evaluation, EGC has concluded that the criteria of 10 CFR 50.92(c) are satisfied and that the proposed UFSAR changes involve no significant hazards consideration.

Page 5 of 7

Attachment I EVALUATION OF PROPOSED CHANGES

8.0 ENVIRONMENTAL CONSIDERATION

In accordance with 10 CFR 50.90, "Application for amendment of license or construction permit,"

Exelon Generation Company, LLC (EGC) is requesting changes to Facility Operating License Nos. DPR-19 and DPR-25, for Dresden Nuclear Power Station (DNPS), Units 2 and 3.

Specifically, the proposed changes will allow EGC to revise the DNPS Updated Final Safety Analysis Report (UFSAR) to allow use of the DNPS Unit 2/3 reactor building crane for a limited duration to lift heavy loads up to a total of 117 tons during removal and installation activities for the reactor shield blocks.

EGC has evaluated these proposed changes against the criteria for identification of licensing and regulatory actions requiring environmental assessment in accordance with 10 CFR 51.21, "Criteria for and identification of licensing and regulatory actions requiring environmental assessments." EGC has determined that these proposed changes meet the criteria for a categorical exclusion set forth in 10 CFR 51.22, "Criterion for categorical exclusion; identification of licensing and regulatory actions eligible for categorical exclusion or otherwise not requiring environmental review," paragraph (c)(9), and as such, has determined that no irreversible consequences exist in accordance with 10 CFR 50.92, "Issuance of amendment," paragraph (b). This determination is based on the fact that these changes are being proposed as an amendment to a license issued pursuant to 10 CFR 50, "Domestic Licensing of Production and Utilization Facilities," which changes a requirement with respect to installation or use of a facility component located within the restricted area, and the amendment meets the following specific criteria:

(1)

The proposed changes -involve no significant hazards consideration.

As demonstrated in Section 7.0, the proposed changes do not involve a significant hazards consideration.

(ii)

There Is no significant change In the types or significant Increase In the amounts of any effluent that may be released offsite.

The proposed changes allow use of the DNPS reactor building crane for a limited duration to lift heavy loads up to 117 tons during removal and installation activities for the reactor shield blocks. There will be no significant increase in the amounts of any effluents released offsite. The proposed changes do not result in an increase in power level, do not increase the production, nor alter the flow path or method of disposal of radioactive waste or byproducts. Therefore, the proposed changes will not affect the types or increase the amounts of any effluents released offsite.

(iii)

There Is no significant increase In Individual or cumulative occupational radiation exposure.

The proposed changes will not result In changes in the configuration of the facility. There will be no change in the level of controls or methodology used for processing of radioactive effluents or handling of solid radioactive waste, nor will the proposal result in any change in the normal radiation levels within the plant. Therefore, there will be no Page 6 of 7

11 Attachment I EVALUATION OF PROPOSED CHANGES increase in individual or cumulative occupational radiation exposure resulting from these changes.

9.0 PRECEDENT The NRC granted a similar license amendment for DNPS in Reference 2.

10.0 IMPACT ON PREVIOUS SUBMITTALS EGC is requesting that the NRC approve the enclosed license amendment request if the Reference 3 amendment request cannot be approved by October 10, 2003. EGC intends to formally withdraw this amendment request if the Reference 3 amendment request is approved by October 10, 2003.

11.0 REFERENCES

1. Letter from Keith R. Jury (Exelon Generation Company, LLC) to U. S. NRC, "Request for License Amendment Related to Heavy Loads Handling," dated September 26, 2002

2. Letter from U. S. NRC to J. L. Skolds (Exelon Generation Company, LLC), "Dresden -

Issuance of Exigent Amendments - Lifting Heavy Loads," dated October 4, 2002

3. Letter from P. R. Simpson (Exelon Generation Company, LLC) to U. S. NRC, "Request for License Amendment Related to Heavy Loads Handling," dated February 26, 2003

4. Letter from J. S. Abel (Commonwealth Edison Company) to U. S. NRC, "Dresden Station Units 2 and 3, Quad Cities Station Units I and 2, Dresden Special Report No. 41, Quad Cities Special Report No. 16, 'Reactor Building Crane and Cask Yoke Assembly Modifications,' AEC Dckt. 50-237, 50-249, 50-254 and 50-265," dated November 8, 1974

5. Letter from J. S. Abel (Commonwealth Edison Company) to U. S. NRC, "Dresden Station Units 2 and 3, Quad Cities Station Units 1 and 2, Dresden Special Report No. 41, Supplement A, Quad Cities Special Report No. 16 - Supplement A, 'Reactor Building Crane and Cask Yoke Assembly Modifications,' NRC Dckts. 50-237, 50-249, 50-254 and 50-265," dated June 3, 1975 Page 7 of 7 PROPOSED REVISIONS TO THE UPDATED FINAL SAFETY ANALYSIS REPORT

DRESDEN - UFSAR Rev. 4 9.1.4.3.2 Reactor Building Overhead Crane The 126-ton capacity reactor building overhead crane main hoist is gle ure of. Within the dual load path, the design criteria are such that all dual elements comply wi e

Specification No. 70 for allowable stresses, except for the hoisting rope which is governed by more stringent job specification criteria. -With several approved exceptions, single element components within the load path (i.e. the crane hoisting system) have been designed to a minimum safety factor of 7.5, based on the ultimate strength of the material. Components critical to crane operation, other than the hoisting system, have been designed to a minimum safety factor of 4.5, based an the ultimate strength of the material. Table 9.14 lists the results of the crane component failure analysis.

The reactor building overhead crane and spent fuel cask yoke assemblies meet the intent of NUREG-0554.

All analyses for handling spent fuel shipping casks, performed relative to the overhead crane handling system loads have been based on the National Lead (NL) 10/24 spent fuel chipping cask which weighs 100 tons (Figure 9.1-18). If larger casks are used, additional analyses will be required to assure safety margins are maintained.

Administrative controls and installed limit switches restrict the path of travel of the crane to a specific controlled area when moving the spent fuel cask. The controls are intended to assure that a controlled path is followed in moving a cask between the shipping axea and the spent fuel pool.

Administrative controls also ensure movement of other heavy loads such as the drywell head, reactor vessel head, and dryer separator assembly is over preapproved pathways.

Technical Specification 3.1014.10 states refueling requirements. Station procedures prohibit movement of heavy loads over the spent fuel pools or open reactor cavity except under Special Procedures.

The crane reeving system does not meet the recommended criteria of Branch Tecinical Position APCSB 9-1 (now incorporated into NUREG-0554) for wire rope safety factors and fleet angles. The purpose of these criteria is to assure a design which minimizes wire rope stress wear and thereby provides maximum assurance of crane safety under all operating and maintenance conditions.

Because the crane reeving system does not meet these recommended criteria, there is a possibility of an accelerated rate of wire rope wear occurring. Accordingly, to compensate in these design areas, a specific program of wire rope inspection and replacement is in place.

The inspection and replacement program assures that the entire length of the wire rope will be maintained as close as practical to original design safety factors at all times. This inspection and replacement program provides an equivalent level of protection to the methods suggested in wire rope safety and crane fleet angle criteria and will assure that accelerated wire rope wear will be detected before crane use.

'Two blocking' is an inadvertently continued hoist which brings the load and head block assemblies into physical contact, thereby preventing further movement of the load block and creating shock loads to the rope and reeving system. A mechanically operated power limit switch in the main hoist motor power circuit on the load side of all hoist motor power circuit controls provides adequate protection 9.1-23

Insert A designated as a single failure proof crane for 10-ton loads. The NRC has approved use of the reactor building overhead crane during power operations to lift a total load up to 116 tons for removal and installation activities for the reactor shield blocks prior to and during Unit 3 refueling outage D3R17 and up to 117 tons for removal and installation activities for the reactor shield blocks prior to and during Unit 2 refueling outage D2R18.

DRESDEN -

UFSAR against two blocking' in the event of a fused contactor in the main hoist control circuitry. This power limit switch will interrupt power to the main hoist motor and cause the holding brakes to set prior to two blocking."

The reactor building refueling floor has been designed for a live load of 1000 b/ 2.

The entire reactor building refueling Bloor (with the exception of the fuel pool and open reactor cavity) is considered a safe load path zone.

A 9-ton load drop has been analyzed. The results show that the refueling foor can survive a drop fim 7 feet without scabbing damage. Procedures limit the 9-ton lift height to a maximum of 7 feet Existing procedural controls limit both the height of a lift to clear obstacles and require the use of the most direct path to laydown areas.

The O bul overhead emeets thestle-ure teiaa in G-0612.

70, e manmum crane eig pusthe weight othe ottom block, divided by the number of parts of rope does not exceed 0% of the manufacturers, published breaking strength.

Te reactor building overhead crane main book has:

A rated load capacity

250,000 lb Block and rope weight

2060D lb Total weight lifted 270,600 lb This weight is supported by 12 parts of wire rope with a published breaking strength of 175,800 pounds.

Totsl weight liftedNumber of parts of rope 270,600 12.8%

(1)

Breaking strength of rope 2

175,800 As canbe seen by Equation 1, this is less than the 20% CMAA-70 requirement.

A detailed analysis of the possibility of horizontal displacement of the cask in the event one of the redundant rope trains fails has been conducted. It has been confirmed that the horizontal load displacement will not exceed 2% inches throughout the critical elevations of lift. At the high point of the lift, with the cask above the operating floor, the static displacement of the load is approximately lb inch with a total static plus dynamic displacement of approximately 1 inch. The total horizontal displacement of the load when the cask is submerged in the spent fuel pool is approxima 2 inches. A larger total horizontal displacement, approximately 9 inches, can occur with the load at its lowest elevation, that is with the load at the grade elevations.: However, it should be noted that the 100-ton cask, which is the heaviest load to be lifted through the equipment -hatchway, is 7 feet, 4 inches in diameter and 7 feet, 10 inches across the cask yoke. The equipment hatchway has a min;unm 20-foot, I-I square opening (See Figure 9.1-20). Local protrusions of ductwork along the vertical path of the cask through the hatchway reduce the cross section to approxiately 19 feet, 6 inches. Since the path of the cask is controlled by limit switches which restrict the position of the cask during lifting to t6 inches from the center line of the hatchway, lateral clearances in excess of 4 feet are available.

S 9,A Insert B The reactor building overhead crane meets the single-failure criteria stated in NUREG-0612 for heavy loads of 110 tons. The NRC has approved use of the reactor building overhead crane during power operations to lift a total load up to 116 tons for removal and installation activities for the reactor shield blocks prior to and during Unit 3 refueling outage D3R17 and up to 117 tons for removal and installation activities for the reactor shield blocks prior to and during Unit 2 refueling outage D2R18.