ML022700281

Request for License Amendment Related to Heavy Loads Handling

ML022700281
Person / Time
Site:	Dresden
Issue date:	09/26/2002
From:	Jury K Exelon Generation Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
-nr, RS-02-168
Download: ML022700281 (18)

Text

Exelkn.

Exelon Generation wwwexeloncorp corn Nuclear 4300 Winfield Road Warrenville, IL 60555 10 CFR 50.59 10 CFR 50.90 RS-02-168 September 26, 2002 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 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, 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 (2)

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 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 (Exelon), LLC, 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 Exelon 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 116 tons for removal and re-installation activities for six reactor shield blocks prior to and during the Unit 3 refueling outage. Reactor shield block removal activities are scheduled to commence on October 7, 2002.

In 1974, Commonwealth Edison (CoinEd) Company, now Exelon, extensively modified the DNPS reactor building crane with the intent of qualifying the crane as single failure-proof for its full rated capacity of 125 tons. In support of a Technical Specifications amendment request to support spent fuel cask handling, we provided information regarding these modifications in References 1 and 2. In this information we stated that the fuel casks used would weigh up to 100 tons with a 10 ton lifting rig.

September 26, 2002 U. S. Nuclear Regulatory Commission Page 2 In a teleconference with the NRC on September 20, 2002, the NRC stated that it considers the DNPS reactor building crane approved as meeting single failure-proof criteria only for loads of up to 110 tons. Subsequently, on September 21, 2002, DNPS determined that the reactor shield blocks, which are moved prior to and during the refueling outage, weigh greater than 110 tons. Since the crane is only approved as single failure-proof for loads of 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 Attachment C, the proposed changes involve no significant hazards consideration.

The weight of the heaviest shield block, including rigging, does not exceed 116 tons. The reactor shield blocks are placed on the refuel floor of the operating unit (i.e., Unit 2). It would be impractical to conduct the refueling outage by placing the reactor shield blocks on the Unit 3 refuel floor. Thus, the requested amendment is needed to prevent a shutdown of Unit 2 to support D3R17. In addition, the requested amendment is needed to allow removal of Unit 3 reactor shield blocks during power operations.

In order to provide a long-term resolution for this issue, Exelon will complete additional analyses and submit a license amendment request related to heavy loads handling. In the interim, Exelon is requesting a one-time license amendment to allow use of the DNPS Unit 2/3 reactor building crane for lifting a total load of up to 116 tons. This will allow DNPS to perform required activities, such as reactor disassembly, for the upcoming Unit 3 refueling outage, D3R17.

Exelon is requesting approval of this amendment by noon on October 7, 2002, on an exigent basis in accordance with 10 CFR 50.91, "Notice for public comment; State consultation,"

paragraph (a)(6). This request meets the criteria of 10 CFR 50.91 (a)(6) because time does not permit the NRC to publish a Federal Register notice allowing 30 days for prior public comment and the requested amendment involves no significant hazards consideration. As described in 10 CFR 50.91 (a)(6)(vi), the exigency could not be avoided by Exelon due to the circumstances described above.

This request is subdivided as follows.

1.

Attachment A gives a description and safety analysis of the proposed changes.

2.

Attachment B provides the proposed revisions to the UFSAR.

3.

Attachment C describes our evaluation performed using the criteria in 10 CFR 50.91(a), "Notice for public comment," paragraph (1), which provides information supporting a finding of no significant hazards consideration using the standards in 10 CFR 50.92, "Issuance of amendment," paragraph (c).

4.

Attachment D provides information supporting an Environmental Assessment.

September 26, 2002 U. S. Nuclear Regulatory Commission Page 3 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 Exelon Quality Assurance Program.

Exelon is notifying the State of Illinois of this request for changes to the operating license by transmitting a copy of this letter and its attachments to the designated State Official.

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

Respe tfully, Keith R. Jury Director - Licensing Mid-West Regional Operating Group Attachments:

Affidavit Attachment A Description and Safety Analysis for Proposed Changes Attachment B Proposed Revisions to the UFSAR Attachment C Information Supporting a Finding of No Significant Hazards Consideration Attachment D Information Supporting an Environmental Assessment cc:

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

STATE OF ILLINOIS COUNTY OF DUPAGE IN THE MATTER OF EXELON GENERATION COMPANY, LLC DRESDEN NUCLEAR POWER STATION, UNITS 2 AND 3

SUBJECT:

)

)

)

)

)

Docket Numbers 50-237 and 50-249 Request for License Amendment Related to Heavy Loads Handling AFFIDAVIT I affirm that the content of this transmittal is true and correct to the best of my knowledge, information, and belief.

Keith R. Jury Director - Licensing Mid-West Regional Operating Group Subscribed and sworn to before me, a Notary Public in and for the State above named, this day of

_________________,206e-7 TMOFFICIA1 QCA1-Nota iP bIk.,

Attachment A DESCRIPTION AND SAFETY ANALYSIS FOR PROPOSED CHANGES A.

SUMMARY

OF THE PROPOSED CHANGES 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 (Exelon), LLC, 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 Exelon 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 116 tons for removal and installation activities for the six reactor shield blocks prior to and during the Unit 3 refueling outage. Reactor shield block removal activities are scheduled to commence at noon on October 7, 2002. The total lifting time of these reactor shield 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 />.

In 1974, Commonwealth Edison (CornEd) Company, now Exelon, extensively modified the DNPS reactor building crane with the intent of qualifying the crane as single failure-proof for the full rated capacity of 125 tons. In support of a Technical Specifications amendment request to support spent fuel cask handling, we provided information regarding these modifications in References 1 and 2. In this information we stated that the fuel casks used would weigh up to 100 tons with a 10 ton lifting rig.

In a teleconference with the NRC on September 20, 2002, the NRC stated that it considers the DNPS reactor building crane approved as meeting single failure-proof criteria only for loads of up to 110 tons. Subsequently, DNPS determined that the reactor shield blocks, which are moved prior to and during the refueling outage, weigh greater than 110 tons. The reactor shield blocks, plus required rigging, weigh less than 116 tons. The reactor shield blocks are placed on the refuel floor of the operating unit (i.e., Unit 2). It would be impractical to conduct the refueling outage by placing the reactor shield blocks on the Unit 3 refuel floor. Thus, the requested amendment is needed to prevent a shutdown of Unit 2 to support D3R17. In addition, the requested amendment is needed to allow removal of Unit 3 reactor shield blocks during power operations.

In order to provide a long-term resolution for this issue, Exelon will complete additional analyses and submit a license amendment request related to heavy loads handling. In the interim, Exelon is requesting a one-time license amendment to allow use of the DNPS Unit 2/3 reactor building crane for lifting a total load of up to 116 tons during power operation.

This will allow DNPS to perform required activities, such as reactor disassembly, for the upcoming refueling outage, D3R17.

Exelon is requesting approval of this amendment on an exigent basis in accordance with 10 CFR 50.91, "Notice for public comment; State consultation," paragraph (a)(6). This request meets the criteria of 10 CFR 50.91 (a)(6) because time does not permit the NRC to publish a Federal Register notice allowing 30 days for prior public comment and the requested amendment involves no significant hazards consideration. As described in 10 CFR 50.91 (a)(6)(vi), the exigency could not be avoided by Exelon due to the circumstances described above.

Page 1 of 5

Attachment A DESCRIPTION AND SAFETY ANALYSIS FOR PROPOSED CHANGES B.

DESCRIPTION OF THE CURRENT REQUIREMENTS 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. The current DNPS licensing basis credits the crane as single failure-proof for handling heavy loads. 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.

C.

BASES FOR THE CURRENT REQUIREMENTS 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 II) 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 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.

In NRC Bulletin 96-02, 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 guidance and requested responses from licensees regarding heavy loads handling.

The DNPS response to Bulletin 96-02 was based on considering the reactor building crane as single failure-proof to 125 tons. This has precluded the need to complete load drop analyses or to restrict movement of heavy loads over safety-related equipment while the reactor is at power.

D.

NEED FOR REVISION OF THE REQUIREMENTS In a teleconference on September 20, 2002, the NRC stated that it has approved the DNPS Unit 2/3 reactor building crane as meeting single failure-proof criteria only for loads of up to Page 2 of 5

Attachment A DESCRIPTION AND SAFETY ANALYSIS FOR PROPOSED CHANGES 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 2/3 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 Attachment C, we have concluded that the proposed changes involve no significant hazards consideration.

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 for general refueling operations because of additional laydown space that is required to be utilized. Sharing of common equipment, such as the refuel bridges, 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 safety due to less restrictive work areas and lower dose rates due to better as low as reasonably achievable (ALARA) practices.

Exelon 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 2), the requested amendment is needed to prevent a shutdown of Unit 2 to support D3R17. In addition, the requested amendment is needed to allow removal of Unit 3 reactor shield blocks during power operations.

E. DESCRIPTION OF THE PROPOSED CHANGES Exelon is proposing to revise the DNPS UFSAR to allow use of the reactor building crane for lifting loads of up to 116 tons to support D3R17. A marked-up copy of the UFSAR has been provided as Attachment B, 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 />.

F. SAFETY ANALYSIS OF THE PROPOSED CHANGES Exelon 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 116 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.

Page 3 of 5

Attachment A DESCRIPTION AND SAFETY ANALYSIS FOR PROPOSED CHANGES An analysis will be performed which will demonstrate that no adverse consequences result from a postulated load drop.

Reactor buildinq 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 1 of Reference 2. 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 116 tons.

A review of References 1 and 2 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 7% when the crane load is restricted to lifting 116 tons. Hence additional margin in the load carrying capacity of critical components will result.

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.

Probability of a Desigqn 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.

Load drop analysis To ensure that the consequences of a load drop are acceptable for the proposed amendment, a load drop analysis is being performed for the proposed movements of the reactor shield plugs. The load drop analysis approach being used meets the intent of the general considerations described in Section 1 of Appendix A to NUREG-0612. The movements of the shield blocks will be restricted to ensure the analysis assumptions are preserved and that no safety-related equipment will be impacted as a result of a postulated load drop. This analysis will be completed prior to lifting the Unit 3 reactor shield blocks.

Page 4 of 5

Attachment A DESCRIPTION AND SAFETY ANALYSIS FOR PROPOSED CHANGES G. IMPACT ON PREVIOUS SUBMITTALS Exelon has reviewed the proposed change and has determined that there is no impact on any previous license amendment request submittals awaiting NRC approval.

H. SCHEDULE REQUIREMENTS We request approval of this license amendment by noon Central Daylight Time on October 7, 2002.

1. REFERENCES

1. 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, 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

2. 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 5 of 5

Attachment B Proposed Revisions to the Updated Final Safety Analysis Report

DRESDEN - UFSAR Rev. 4 9.1.4.3.2 Reactor Building Overhead Crane The 125-ton capacity reactor building overhead crane m. ain hoist is*

gle ure of. Within the dual load path, the design criteria are such that all dual elements comply with the CMA 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 on the ultimate strength of the material. Table 9.1-3 lists the results of the crane component failure analysis.

The reactor building overhead crane and spent fuel cask yoke assemblies mee 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 shipping 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 area 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 3.10/4.10 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 Technical 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 110-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.

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 lb/ft2.

The entire reactor building refueling floor (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 floor can survive a drop from 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.

rThe rbuil overhead emeets theviftee-fhi--

m G-0612.

requir--70, the maximum crane loa weig plus the weight or the bottom block, divided by the number of parts of rope does not exceed 1 0% of the manufacturefs published breaking strength.

The reactor building overhead crane main hook has:

A rated load capacity

=

250,000 lb Block and rope weight

=

20,500 lb Total weight lifted

= 270,500 lb This weight is supported by 12 parts of wire rope with a published breaking strength of 175,800 pounds.

Total weight lifted/Number of parts of rope 270,500

= 12.8%

(1)

Breaking strength of rope 12 x 175,800 As can be 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 21 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

.3h 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 approximately 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 minimum 20-foot, 1-inch 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 approximately 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.

9.1-24

Insert B The reactor building overhead crane meets the single-failure criteria stated in NUREG-0612 for heavy loads of 11 0-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 D3R1 7.

Attachment C INFORMATION SUPPORTING A FINDING OF 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 (Exelon), LLC, is requesting changes to Facility Operating License Nos. DPR-1 9 and DPR-25, for Dresden Nuclear Power Station (DNPS), Units 2 and 3.

Specifically, the proposed changes will allow Exelon 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 116 tons for removal and re-installation activities for six reactor shield blocks prior to and during the Unit 3 refueling outage. Reactor shield block removal activities are scheduled to commence on October 7, 2002.

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

The proposed changes will allow use of the reactor building crane at DNPS during power operations to lift heavy loads up to 116 tons for removal and installation activities for the reactor shield blocks prior to and during the Unit 3 refueling outage (i.e., D3R17). Reactor shield block removal activities are scheduled to commence on October 7, 2002. The reactor building crane has additional margin for a total lifted load of 116 tons with single failure-proof features if a Design Basis Earthquake (DBE) is not assumed. Exelon 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 previously evaluated is not increased since the capacity of the reactor building crane exceeds the weight of the reactor shield blocks. A load drop analysis will also be performed to demonstrate that dropping a 116 ton load will not result in a significant increase in the consequences of an accident previously evaluated.

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

Page 1 of 2

Attachment C INFORMATION SUPPORTING A FINDING OF NO SIGNIFICANT HAZARDS CONSIDERATION 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 116 tons during removal and installation activities for the reactor shield blocks. The reactor building crane has additional margin for a lifted load of 116 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 116 tons. Existing safety margins are enhanced when lifting loads up to 116 tons if a DBE is not assumed, and Exelon 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, Exelon 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 2 of 2

Attachment D INFORMATION SUPPORTING AN ENVIRONMENTAL ASSESSMENT In accordance with 10 CFR 50.90, "Application for amendment of license or construction permit," Exelon Generation Company (Exelon), LLC, 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 Exelon 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 116 tons during removal and installation activities for the reactor shield blocks.

Exelon 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." Exelon 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:

(I)

The proposed changes involve no significant hazards consideration.

As demonstrated in Attachment C, 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 116 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.

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Attachment D INFORMATION SUPPORTING AN ENVIRONMENTAL ASSESSMENT (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 increase in individual or cumulative occupational radiation exposure resulting from these changes.

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