ML19344A908

Forwards Response to 780517 Ltr Re Control of Heavy Loads on Spent Fuel.Maint Procedures M-0305,M-0306,M-0913 & Drawings Encl

ML19344A908
Person / Time
Site:	Saint Lucie, Turkey Point
Issue date:	10/04/1978
From:	Robert E. Uhrig FLORIDA POWER & LIGHT CO.
To:	Stello V Office of Nuclear Reactor Regulation
Shared Package
ML19344A909	List: ML19344A908 ML19344A910
References
REF-GTECI-A-36, REF-GTECI-SF, TASK-A-36, TASK-OR L-78-326, NUDOCS 8008220581
Download: ML19344A908 (45)

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5083 FLORIDA POWER & LIGHT COMPANY October 4, 1978 L-78-326 THIS DOCUMENT CONTAINS w; POOR QUAUTY PAGES Director of Nuclear Reactor Eegulation Atteation: Mr. Victor Stello, Director Division of Cperating Reactors U. S. Nuclear Regulatory Com:aission Washington, D. C. 20535

Dear Mr. Stello:

Re: St. Lucie Unit 1 Docket No. 50-335 Control of Heavy Loads Responses to the NRC's information request of May, 17, 1978, regarding

" Control of Heavy Loads Near Spent Fuel" are attached. The Staff'c letter Ir.akes mention of the passibility of radiat ion releases in excees of 10 CFR Part 100 guidelines as the basis for its requent. It should be noted at this point that the licensing requirements for postulated radiological releases to the environment and potential radiation expasures to inplant personnel resulting from recidents involving spent fuel were reviewed and approved by the Staff during the licensing process for St. Lucie Unit No. 1. These are reflected in the unit's Technical Specifications. Conservative analymes of postulated radiological releases demonstrate that the resultant off-cite doses are well below 10 CFR 100 guidelines. The facility emergency plan and procedures administrative 1y provide appropriate assurance that doses to inplant personnel are acceptably low and will not impact post-accident unit cafety requirements.

Regarding the Staf f's statement on improving margins, it is further noted that in its " Summary of Meeting to Discuss Margin in Technical Specifications", dated May 13, 1976, f ollowing a meeting in response to FPL letter dated March 11, 1976, the Staf f stated ". . .that it is not it.'s (sic) intent to impose additional safety nargins in the Technical Specifica: ions beyond those considered to exist in the staff's safety evaluation. But, that it is one of the functions of the Technical Edolfzz06fs' PEOPLE. . SERVING PEOPLE

Director of Nuclear Reactor Regulation Page Two Specifications to assure that the safety margins considered to exist in the staff's safety evaluation do, in fact, continue to exist throughout the life of the facility". Since the St. Lucie Unit No. 1 facility Technical Specifications currently provide this assurance, further revision is neither indicated nor required.

Very truly yours, Robe t E. Uhrig we President REU/ MAS /cpc Attachment cc: Mr. James P. O'Reilly, Region II Harold.F. Reis, Esquire i

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ATTACHMENT Re St. Lucie Unit 1 Docket No. 50-335 Control of Heavy Loads Question 1: Provide a diagram which illustrates the physical relation between the reactor core, the fuel transfer canal, the spent fuel storage pool and the set down, receiving or

. storage areas for any heavy loads moved on the refueling floor.

Response la The attached drawings, 8770-G-073 and 8770-G-065, provide the general' arrangement of the Reactor Building and the Fuel Handling Building and the laydown areas for heavy loads that are required to be moved over the reactor (during refueling), or the spent fuel storage pool.

(These drawings are also shown in the St. Lucie Unit 1 FSAR, Figures 1.2-8 and 1.2-18, respectively.)

'v Question 2: Provide a list of all objects that are required to be moved over the reactor core (during refueling), or the spent fuel storage pool. For each object listed, provide its approximate weight and size, a diagram of the movement path utilizedy (including carrying weight) and the frequency of movement.

Response 2: The following table delineates all objects required to be moved over the reactor core (during re fueling) after the reactor vessel head is removed from then vessel:

NOMINAL ST. LUCIE DATA UNIT 1 Approximate Approximate Approximate Approximate Carryigg Anticipatgd Size Weicht Height Frequency R3tctor Vessel Head 205" dia. x 8' 158,400 lbs. 30 ft. 2 Rerctor Vessel 150" sq. x 14' 128,800 lbs. 11 ft. 2 Upper Internals (with lifting rig)

Upper Internal Lifting Rig 200" dia. x 238" 14,000 lbs. 20 ft. 2 Fu21 Handling Tool 3" dia. x 35' 600 lbs. 25 ft. 520 Sp;nt Fuel Assemblies 8" sq. x 157" 1,579 lbs 4 ft. 60 New Fuel Assemblies 8" sq. x 157" 1,579 lbs. 4 ft. 60 1

Loads are moved from their point of origin to the laydown area as shown on drawings provided in response to Question 1.

The objects required to be moved over the spent fuel storage pool are; ruol Hand?ing Tool 42' Long 100 lbs 5 f t. 60 Spent Fuel Shipping Cask 50" dia. x 200" 50,000 lbs. 42 ft. As Required Spent Fuel Assemblies 8" sq. x 157" 1,280 lbs. 5 ft. 60 Shipping Cask Cover 25-1/2" dia. x 8-3/8" 750 lbs. 5 ft. As Required Approximate height above Reactor Vessel Flange or Spent Fuel Racks 3

Frequency per refueling or per year

Question 3: What are the dimensions and weights of the spent fuel casks that are or will be used at your facility?

Response 3: The following is a table of spent fuel casks which have been used or are currently available for use at St. Lucie No. 1.

Cask Height ' Base Diameter Weight Shape National Lead Industry *193" 33" 25 tons right circular a NLI 1/2 LWT (1) cylinder J

Nuclear Assurance Corp. *200" **50" 25 tons right circular NAC - 1 (2) (NFS-4) cylinder

• Shipping lid removed during fuel handling operations.

• • Base of cask; body diameter N39.2".

(1) Docket No. 71-9010 (2) Docket No. 71-6698

A Question 4: Identify any heavy load or cask drop analyses performed to date for your facility. Provide a copy of all such analyses not previously submitted to the NRC staff.

f Response 4: Section 9.1.4.3 of the FSAR provides two postulated cask drop accidents for St. Lucie. A vertical drop is analyzed to determine if the leaktight barrier of the spant fuel pool can be breached. As per our response to Question 6, the leaktight integrity of the Spent Puel Pool is maintained for this case.

A tipped cask drop was also considered. The radiological consequences of the tipped cask drop were provided in FSAR Section 15.4.3 and were subsequently updated by the " Spent Fuel Storage Facility Modification Safety Analysis Report" to accommodate the increased spent fuel storage capability. Reportable Occurrence 335-78-25 dated August 7, 1978, notified the Staff of an error in the original FSAR cask drop analysis. A reanalysis, wherein it is conservatively assumed that all spent fuel can be struck by the carx drop, has been completed. The appropriate Technical Specification, bases, and associated safety evaluation were transmitted with the proposed Technical Specification amendment by letter dated October 2, 1978 (L-78-317).

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Qu:stion 5: Identify any heavy loads that are carried over equipment required for the safe shutdown of a plant that is operating at the time the load- is moved. Identify what equipment could be affected in the event of-a heavy load handling accident (piping, cabling, pumps, etc.) and discuss the feasibility of sucn an accident affecting this equ.pment. Describe the basis for your conclusions.

Re:ponse: St. Lucie Unit 1 is an operating plant and as such no movement of heavy loads over equipment required for safe plant shutdown at the time the load is moved would be expected within con-tainment or other indoor areas (RAB, D/G building) containing safe shutdown related equipment. All roads in the plant area for Unit 1 are designed for H-20 loading in accordance with the American Association of State Highway Officials (AASHO). All pipes and duce banks under on-site roads are designed to resist H-20 wheel Ic.:ds.

Roads along the route used by the fuel handling cask transporter are designed for the transporter wheel loads. All duct banks and pipes under these roads are designed to withstand the loads from the fuel handling cask transporter wheels. There is no safe

-shutdown equipment along .the fuel handling cask transporter route.

During transportation of Unit 2 NSSS components, the transport path from the NSSS component laydown area to the construction batch in the Unit 2 Containment Building will be provided with cribbing or other load distribution devices as required to l

. assure proper load distribution on plant roads. All buried  !

facilities in the path of these heavy equipment loads are designed l to accommodate these load distributions. NSSS transport vehicle )

- tumiirg radius and speed will be administratively controlled, thus,

'8 the overturning of the transport vehicle is. highly unlikely.

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Quastion 6: If heavy loads are required to be carried over the spent fuel storage pool or fuel transfer canal at your facility, discuss the feasibility of a handling accident which could result in water leakage severe enough to uncover the spent fuel. Describe the basis for your conclusions. .

Re:ponse: Section 9.1.4.3 of the FSAR provides an analysis for a 25 ton cask drop from the maximum lift height of 58 feet. The analyzed target area for this cask drop analysis was the cask storage area in the northeast corner of the pool. The results of this analysis demonstrate- that spect fuel pool leaktight integrity is maintained. Additionally it should be noted that by design the wall between the cask storage area and the spent fuel pool precludes uncovering of the spent fuel even if leakage in the cask storage area is postulated. It should be further noted that in the event that leakage is postulated there are numerous alternative makeup water sources available to the spent fuel -

pool as discussed in FSAR Section 9.1.3. _.

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Question 7: Describe any design features of your facility which affect the potential for a heavy load handling accident involving spent fuel,

. e.g., utilization of a single failure-proof crane.

The St. Lucie Unit 1 fuel handling system is designed by means of ROsponse 7:

interlocks, travel limit devices, and other protective devices to minimize the probability of malfunction or operator initiated actions that could cause spent fuel damage. As discussed in Section 9.1.4 of the FSAR.

a) The fuel cask handling crane does not allow the cask to travel over spent fuel because of the geometrical design of the crane runway, the physical design of the building, and the travel limit switch interlock circuitry.

b) The crane is equipped with a two cable system with an associated safety factor of 6.9. In the unlikely event of a cable failure the safety factor is reduced by only one-half.

c) The hoist motors are torque limited to approximately 200%; when this torque is reached the electrical circuit is tripped.

d) Diverse and redundant upper limit switches preclude block engagement.

e) With loss of electrical power both stopping / holding brakes automatically set. They are capable of stopping a 105 ton load within 1" assuming maximum lowering speed. In the unlikely evant of a brake failure, the remaining brake can stop this load in 2".

Qurstion 8:- Provide copies of all procedures currently in effect at your facility for the movement of heavy loads over the reactor core during refueling, the spent fuel storage pool, or equipment required for the safe shutdown of a plant that is operating at the time the move occurs.

Response 8: Procedures currently in effect at St. Lucie for movement of heavy loads near spent fuel are attached:

M-0305 Reactor Vessel Closure Head - Removal and Installation M-0306 Reactor Vessel Internals - Removal and Installation M-0913 General Operating Procedure for Handling and Loading of the NFS-4 cask

t Qusation 9: Discuss the degree to which your facility complies with the eight (8) regulatory positions delineated in Regulatory Guide 1.13 (Revision 1, December, 1975) ,regarding Spent Fuel Storage Facility Design Basis.

Re:ponse: St. Lucie Unit #1 meets "the intent of the regulatory positions of Safety Guide 1.13 (3/10/71) as presented in the FSAR sections referenced below: .

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That .dollowing is a listing of the FSAR sections where a repsonse to each position may be found:

Regulatory Position 1: 3.7.5, 3.8, 9.1.2.3 .

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Regulatory Position 2: - 3.'2, 3.8, 9.1.2.3 Regulatory Position 3: 3.7, 9.1.2.3

. Regulatory Position 4: 7.6.1.2, 9.4.6, 12.2.2.5, 15.4.3 Regulatory Position 5:

9.1.2.3, 9.1.4.3*, 15.4.3

Regulatory Position 6

9.1.3.4 Regulatory Position 7: 9.1.2.5, 9.1.3.6, 12.1.4 Regulatory Position 8: 9.1.3.4 .

• As modified by the Spent Fuel Storage Facility Modification Safety Analysis Report transmitted via L-77-273, dated 3/31/77. As described therein, no other modifications were required in the above. Referenced FSAR/ Regulatory Guide 1.13 positions are applicable.

A review of the changes in Revision 1 shows that the facility design is capable of meeting the intent of the 12/75 version of the s

Regulatory Guide as well.

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