ML20148G751

Application for Amend to License DPR-3,adding Shipping Cask Liners,Vol Reduction Equipment,Shipping Casks Not for Spent Fuel & Spent Fuel Assembly Nondestructive Test Equipment to Tech Spec 3.9.7.Fee Paid

ML20148G751
Person / Time
Site:	Yankee Rowe
Issue date:	01/15/1988
From:	Heider L YANKEE ATOMIC ELECTRIC CO.
To:	Fairtile M NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM), Office of Nuclear Reactor Regulation
Shared Package
ML20148G758	List: ML20148G751 ML20148K167
References
FYR-88-010, PC-212, NUDOCS 8801270025
Download: ML20148G751 (7)

Text

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Tshphons (617) 872-8100 TWX 710-380 7819 YANKEE ATOMIC ELECTRIC COMPANY y

% y 1671 Worcester Road, Framingham, Massachusetts 01701 January 15, 1988 FYR 88-010, PC-212 United States Nuclear Regulatory Commission Document Control Desk Washington, DC 20555 Attention: Mr. Morton B. Fairtile, Project Manager Project Directorate I-3 Division of Reactor Projects I/II

References:

(a) License No. DPR-3 (Docket No. 50-29)

(b) YAEC Letter to NRC, dated February 12, 1982 (FYR 82-19)

(c) YAEC Letter to NRC, dated November 4, 1983 (FYR 83-93)

(d) NRC Letter to YAEC, dated February 19, 1985 (NYR 85-37)

Subj ect: Crane Travel - Spent Fuel Pit

Dear Sir:

Pursuant to Section 50.59 of the Commission's Rules and Regulations, Yankee Atomic Electric Company hereby proposes the following modification to Appendix A of the Operating License.

PROPOSED CHANGE Reference is made to the Technical Specifications of the Yankee Nuclear Power Station's Operating License No. DPR-3. We propose to modify the license as follows:

1. Add the following exceptions to Technical Specification 3.9.7:

g. Shipping cask liners.

h. Volume reduction equipment.

i. Shipping casks: 35 tons maximum gross weight - not for spent fuel, J. Spent fuel assembly nondestructive test equipment.

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United States Nuclear Regulatory Commission January 15, 1988 Attention: Mr. Morton B. Fairtile- Page 2 FYR 88-010

2. Revise Technical Specification 4.9.7.1 Exceptions a and b, to read:

a. Spent Fuel Pit Building hatches, the spent fuel inspection stand, the fuel. handling equipment, spent fuel racks, the temporary gate, the shielding panels, the spent fuel assembly nondestructive test equipment, the volume reduction equipment, the shipping cask liners, and the shipping casks (35 tons

-maximum gross weight - not for spent fuel) may travel over the spent fuel pit in accordance with approved written procedures;

b. The spent fuel inspection stand, the temporary gate, the shielding panels, the spent fuel assembly nondestructive test equipment, and the cask hatch cover shall be prevented from traveling over fuel assemblies in the spent fuel pit by administrative control. The volume reduction equipment, the shipping cask liners, and the shipping casks shall be prevented from traveling over fuel assemblies in the spent fuel pit by administrative control and by the steel framing at the southern edge of the spant fuel pit superstructure roof opening which is also at the southern end of the safe load path; and Revise Technical Specification Bases 3/4.9.7, second paragraph, second and third sentences to read:

The restriction on movement of the spent fuel inspection stand, the temporary gate, the shielding panels, the spent fuel assembly nondestructive test equipment, the cask hatch cover, the volume reduction equipment, the shipping cask liners, and the shipping casks over spent fuel ensures that these items cannot be dropped on spent fuel. Dropping any one of these items from its maximum height will not result in loss of integrity of the fuel pool floor.

Delete Technical Specification Bases 3/4.9.7, second paragraph, last sentence.

REASON FOR CHANGE Current Technical Specification 3.9.7 allows only certain identified exceptions to the 900-pound fuel assembly weight limit for travel over the Spent Fuel Pit (SFP). An inventory of activated nonfuel components, predominately control rods and follower sections, has accumulated in the SFP.

The need for additional fuel storage space dictates a SFP cleanout prior to the 1988 refueling outage. In order to remove the control rods, it will be necessary for volume reduction equipment weighing approximately 10,000 pounds, shipping cask liners with gross weight less than 2,000 pounds, and shipping casks with max 3 mum gross weight equal to 70,000 pounds to travel over the northern end of the SFP. In order to perform nondestructive testing of spent fuel assemblies, it will be necessary to periodically lift test equipment weighing less than 2,000 pounds into the northern end of the SFP.

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' United States Nuclear Regulatory Commission January 15, 1988 Attention: Mr. Morton B. Fairtile Page 3 FYR 88-010 BASIS FOR CHANGE A. Removal of Control Rods

- 1. Safe Load Path The SFP is a reinforced concrete structure 37 feet deep with a. minimum of 14 feet below grade. The SFP Building enclosure is a steel framed superstructure- over the SFP.

The cask opening in the SFP Building enclosure was enlarged in 1982 to allow side entry of the cask and related equipment into the SFP. With the new opening the maximum height of a cask lift has been reduced from 52' to 38' above the floor of the SFP. Also, the steel framing around the SFP superstructure roof opening prevents movement of the cask over the spent fuel.

The safe load path for the casks, cask liners, and volume reduction equipment, as shown on the attached sketch, involves simple vertical and horizontal movements (not performed simultaneously) during which the cask is totally visible at all times by the signal man. The primary horizontal motion will be north-south bridge motion. Slight east-west trolley motion may be required for adjustment while moving through the cask opening and just prior to setting down on the pit floor. Vertical motion will take place while lowering and lifting inside and outside the SFP.

2. Yard Area Crane The crane to be used for this operation is the yard area crane. It has a capacity rating of 75 tons with a safety factor of 5 against ultimate strength. For the 35-ton cask, the working safety factor against ultimate strength of the crane is 10.7.

The yard area crane has a long history of successful operation, including 34 fuel shipments with a 75-ton spent fuel shipping cask.

These lifts were nearly identical in nature to those proposed for shipping control rods, except that the cask for control rods is much lighter and the side entry modification to the SFP Building enclosure has reduced the lift height from 52' to 38'.

The design of the yard area crane complies with the guidelines of the Crane Manufacturers Association of America (CMAA) Specification No. 70, and Chapter 2-1 of ANSI B30.2-1976, as outlined in Reference (b).

The main hoist rope of the crane has a safety factor against ultimate strength of 5.2 for the 75-ton rated load. For the 35-ton cask, the working safety factor against ultimate strength of the main hoist rope is 11.1. Therefore, safety factors for this lift are greater than 10 which is consistent with the defense-in-dopth approach recommended in Section 5.1 of NUREG-0612.

b . .

United States Nuclear Regulatory Commission January 15, 1988 Attention: Mr. Morton B. Fairtile Page 4 FYR 88-010 Both the main and auxiliary hoists of the yard area crane are equipped with screw-type limit switches to limit travel in both the raising and lowering modes. In addition, redundant limit switches have been installed to provide an additional margin of safety.

The crane support structure is a structural steel braced frame. It was designed in accordance with the AISC code and has a safety factor against yield of 1.67. For the 35-ton cask, the working safety factor against yielding of the support structure is 2.9. The support structure has redundant framing members so that gross failure would be extremely unlikely due to yielding at one location because the forces would be redistributed to another load path.

3. Plant Procedures Plant procedures have been revised to include the recommendations identified in NUREG-0612, Section 5.1.1(2), and to include our commitments identified in References (c) and (d). Per these procedures, inspections, maintenance, and operational testing will be performed prior to the cask lifts. Also, a load test equal to 150% of the maximum load to be lifted will be performed. After the load test, nondestructive examination of critical stress components including the crano hook will be performed.

Special precautions will be taken during lifting of the cask over the SFP. Plant procedures require that the crane operator has been trained and qualified in accordance with Chapter 2-3 of ANSI B30.2-1976.

Supervision of the cask lift will be performed by a Senior Reactor Operator. A cask lift procedure will be prepared prior to this task.

4. Cask Lifting Device The cask lifting fixture conforms to the requirements of NUREG-0612, Section 5.1.1(4). In addition to these requirements, the lifting fixture will have dual load paths or a minimum factor safety of 10,

5. Cask Drop Analysis A cask drop analysis has been performed which assumes that the cask is dropped in an orientation which results in the most severe consequences to the SFP. The maximum gross weight of the loaded cask, 37 tons including water in the cask, was assumed to be dropped the maximum height, 38', on the concrete floor slab of the SFP. The effect of the drag forces caused by the water in the SFP was included. EPRI's ABAQUS-EPGEN structural analysis code was used for the analysis of the concrete floor slab. The cask, stainless steel liner, concrete slab, and soil were modeled using finite elements. The model for the concrete floor slab accounts for tensile cracking, compressive plasticity, strain sof tening, and crushing. A normalized stress-strain curve, representative of the concrete, was used to obtain properties of the concrete for all stages of loading.

United States Nucicar Regulatory Commission January 15, 1988 Attention: Mr. Morton B. Fairtile Page 5 FY3 88-010 The reruits of the analysis show that there is minimal cracking of the concrete slab but the steel liner would not fail. Therefore, leakage would not occur in the unlikely event of a cask drop.

If the liner were breeched, however, calculations accounting for the soil permeability indicated that the maximum leakage would be less than 20 gallons per minute (gpm) which is well below the makeup rate for the SFP.

B. Placement of Spent Fuel Assembly Test Equipment in The SFP

1. Plant Procedures Lifting of the spent fuel assembly test equipment will be performed using the same existing procedures for the yard area crane as are used for the lifting of the cask and related equipment. A specific lif t procedure will be prepared for this task. The load path over the SFP will also be identical to that used for control rod removal.

2. Drop Analysis The total maximum weight of this test equipment is less than 2,000 pounds. The consequences of an accidental drop are enveloped by the cask drop analysis described above.

SAFETY CONSIDERATION This change is requested in order to remove activated nonfuel components from the SFP and to allow installation and removal of spent fuel assembly test equipment at Yankee Nuclear Power Station. As such, this proposed change would not:

1. Involve a significant increase in the probability or consequences of an accident previously evaluated. The large safety factors associated with these specific tasks ensure that the probability of an accident will not be significantly increased. The results of the cask drop analysis indicating no leakage ensures that the consequences of an accident will not be significantly increased.

Therefore, this change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

2. Create the possibility of a new or different kind of accident from any previously evaluated. The use of specific lift procedures in conjunction with the revised plant procedures and the safe load path ensures that this change does not create the possibility of a new or different kind of accident from any previously evaluated.

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3. Involve a significant reduction in a margin of safety. The use of I

procedures involving inspection, operational and load testing, crane operator training requirements, and precautions ensures that this change does not involve a significant reduction in margin of safety. ,

United States Nuclear Regulatory Commission January 15.-1988 Attention: Mr. Morton B. Fairtile Page 6 FYR 88-010 Based on the considerations contained herein, it is concluded that there is reasonable assurance that operation of the Yankee plant, consistent with the proposed Technical Specifications, will not endanger the health and safety of the public. This proposed change has been reviewed by the Nuclear Safety Audit and Review Committee.

FEE An application fee of $150.00 is enclosed in accordance with 10CFR170.21.

SCHEDULE OF CHANGE These changes to the Yankee Technical Specifications will be implemented upon Commission approval. In accordance with our plans for the Cycle 20 refueling, we request approval of this submittal prior to April 15, 1988.

We trust that you will find this submittal satisfactory; however, should you desire additional information, please contact us.

Very truly yours, YANKEE ATOMIC ELECTRIC COMPANY

$M L. H. Heider Vice President / Manager of Operations I

LHH/3.91 Attachment Enclosure COMMONWEALTH OF MASSACHUSETTS)

)ss MIDDLESEX COUNTY )

Then personally appeared before me, L. H. Heider, who, being duly sworn, did state that he is Vice President and Manager of Operations of Yankee Atomic Electric Company, that he is duly authorized to execute and file the foregoing document in the name and on the behalf of Yankee Atomic Electric Company and that the statements therein are true to the best of his knowledge and belief.

Robert H. Groce Notary Public My Commission Expires August 29, 1991 cc: USNRC Region I USNRC Resident Inspector YNPS

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