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Forwards Results of Bechtel Study of Postulated Cask Drop Accident.Indicates That Cask Dropped Into Spent Fuel Pool Caused Leak Greater than Makeup Capacity of Backup Sys. Detailed Mods to Equipment & Procedures Necessary
	ML19305B749
Person / Time
Site:	Big Rock Point File:Consumers Energy icon.png
Issue date:	07/01/1974
From:	Sewell R; CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.)
To:	Goller K; US ATOMIC ENERGY COMMISSION (AEC)
Shared Package
ML19305B745	List: ML19305B744; ML19305B748; ML19305B749; ML19305B750; ML19305B751; ML19305B758;
References
	NUDOCS 8003200246
	Download: ML19305B749 (77)
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{{#Wiki_filter:, p q.O ./,. -- CODSumerS ).m i. } F0W8r W. / Campany N../ ~ ~.. - ~ rs D YJ [ ....o ~....,.............~......_..................c............e July 1, 1974 Mr. Karl R. Goller Re: Docket 50-155 Assistant Director for Operating Reactors License CFR-6 Directorate of Licensing US Atomic F.nergy Cc d ssion Washington, DC 205h5

Dear Mr. Goller:

By letter dated February h,197h, it was requested that we pro-vide an analysis and other relevant infor=ation needed to deter =ine the possible damage in the event of a cask drop caused by a syste= failure and whether design or procedural =odifications vould be appropriate to reduce the probability of occurrence for our Big Rock Point Plant. This study gm of a postulated ca_h drop accident was perfor=ed by Bechtel Corporation at Consu=ers Power Co. pany's request. The results of this study are con-tained in Attach =ent A. The results of this analysis reveal that a cask being dropped into the spent fuel pool could result in a leak in the fuel pool which might be larger than the =akeup capability of the backup syste=s. The decay heat being generated inside each fuel pin vould eventually =elt the fuel pin cladding. The resulting site boundary dose with = ore than ten (10) bundles in the fuel pool vould exceed those limits as specified in 10 CFR 100. In order to eliminate this potential cask drop accident as well as other less severe cask drop accidents fro = occurring, detailed =odifi-catiens to the present procedures, equip =ent and structures are proposed. These proposed modifications were based en a =ultilevel concept of syste= safety and include: 1. Modifications of the crane coupled with' detailed handling procedures to eli=inate the possibility of a cask drop. 2. Tamit-svitches and specific cask =ovement paths minimize possible endangered areas and restrict the =axi=u:n drop heights. 3 Potential i= pact areas inside those restricted areas are \\ i j l protected if necessary and the fuel bundles relocated to provide a further safety factor. 8003200@h +

s o 9 Mr. Karl R. Goller 2 Docket 50-155, License DPR-6 July 1, 1974 ( In addition, the low usage factor of the crane for those weights approaching the crane capacity help to minimize stresses in the crane systems and, therefore, significantly reduce any probabi.11ty of failure. The 60-ton refueling cask is expected to be utilized 4 two to three times per year. The procedures as described above are being written and imple-mented while the :nodification to the crane and the fabrication and instal-lation of the impact pad vill be initiated upon your approval of these recenmended rclifications. It is estimated that these =odifications can be ccmpleted within one year of receipt of your approval. Yours very truly, R. B. Sewell (Signed) JSR/ nap R. B. Sewell Nuclear Licensing Administrator CC; JGKeppler, USAEC

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s \\ 4 ATTACHMENT A )N / D i 1 l 1 i 1 h't ) \\

s TABLE OF CONTENTS Section Page 1-1

1.0 INTRODUCTION

? 2.0

SUMMARY

........................................... 2-1 3.0 D E S C RIP T IO N....................................... 3-1 3.1 Casks............................................. 3-1 3.2 Spent Fuel Handling Cranes and Hoists............. 3-2 3.3 Handling Procedures............................... 3-5 3.4 Structures 3-8 3-11 3.5 Safety Related Equipment 3.6 Equipment Tests................................... 3-14 x 4.0 SPENT FUEL HANDLING CRANES AND HOISTS INVESTIGATION 4-1 4.1 Structural........................................ 4-1 4.2 Mechanical 4-1 4.3 Controls and Electrical........................... 4-5 b: .rg 5.0 CASK DROP ACCIDENTS 5-1 5.1 Spent Fuel Shipping Cask.......................... 5-1 5.2 Fuel Transfer Cask.................'.............. 5-3 5.3 Cobalt Cask...................................... 5-4 5.4 Fuel Rod (TREAT II) Cask......................... 5-4 6.0 EFFECTS ON STRUCTURAL INTEGRITY 6-1 7.0 RESPONSE TO LOSS OF SAFETY-RELATED EQUIPMENT 7-1 8.0 ENVIRONMENTAL EFFECTS 8-1 9.0 RECOMMENDED MODIFICATIONS...........'.............. 9-1 APPENDI/ES 1 i 7-A Sample Calculations............................... A-1 g 1 , sf ' B Tests of Equipment and Components Used for ) Handling Casks u................................... B-1 ) 1 ~ l I L i

__7_ .._1 LIST OF FIGURES h, s No. Title 3-1 Cutaway Of Containment Vessel 3-2 Spent Fuel Storage Racks A and B, Details 3-3 Spent Fuel Storage Rack C, Details 3-4 Plan Below Elevation 598 '-0" 3-5 Plan Below Elevation 584'-0" 5-1 Plan Of Postulated Cask Drop Accidents 5-2 North-South Section Through Reactor Building 5-3 East-West Section Through Reactor Building 9-1 Plan Of Spent Fuel Storage Pool And Proposed Modifications 9-2 Proposed Cask Routing 9-3 Proposed Impact Pads LIST OF TABLES b "Q ) No. Title 3-1 Structural Description 6-1 Spent Fuel Shipping Cask, Postulated Cask Drop Accidents 6-2 Fuel Transfer Cask, Postulated Cask Drop Accidents 6-3 Cobalt Cask, Postulated Cask Drop Accidents ~ 6-4 Fuel Rod (TREAT II) Cask, Postulated Cask Drop Accidents 6-5 Cask Drops On Fuel Racks 7-1 Correspondence Between Postulated Accidents And Plant Safety Equipment 8-1 Site Boundary Dose Per Crushed Fuel Bundle 8-2 Site Boundary Dose Per Violated Fuel Bundle WP' - - ~ ~ - - - -.-. --

1.0 INTRODUCTION

) This report contains analyses and evaluation of the conse-s quences of postulated fuel cask drop accidents in the reactor building at Big Rock Point Nuclear Power Plant. Estimates of damage to structures and safety related equipment and the effects on the environment are included. The dropping of the casks is assumed to be the result of a hypothetical failure in the crane system or handling devices. i An assessment of the present crane system, handling devices and handling procedures is included, along with recommended changes and improvements. t [ ?' The casks considered in this investigat n are the fuel ship-ping cask, e cobalt cask, the fuel ro7 ' 7(TREAT II) ' MT cask, and 7 the fuel nsfer cask. )(r^T The casks are assumed to be dropped over the spent fuel pool ~ (and spent fuel storage racks) in the cask storage area, decon-tamination area, and various locations in the building on routes along.which the casks are moved. This investigation includes the effects of the postulated cask drops on the integrity of impacted structures and equipment and an evaluation of the capability of floors subject to cask drop to protect safety related equipment and systems located beneath the floors. In evaluating postulated cask drop accidents, the present handling system and current handling procedures were used. l A sw= mary of this investigation and recommended modifications l to minimize the consequences or occurrence of a cask drop acci-l!,~' dent are contained in Section 2. l Jr ' 1-1

- 2.0

SUMMARY

) The presene cask handling system and operating procedures have s been reviewed and modified to provide for an increased margin of safety. The present handling equipment includes: a. 75 ton single leg gantry crane, b. 75 ton spent fuel handling crane. c. Slings, yokes and handling attachments. A description of the design and safety features of this equip-ment is contained in Sections 3.2 and 4.0. Handling procedures are summarized in Section 3.3. For this investigation, four casks were assumed to be involved in postulated cask drop acci-dents. Details of the spent fuel shipping cask, fuel transfer cask, cobalt shipping cask and the TREAT II cask can be found in Section 3.1. A structural analysis of the containment building revealed nine ,3[ representative areas where cask drops might result in damage.

L These postulated occurrences and their locations are sc=marized in Section 5.

The resulting effects on the surrounding struc-tures, equipment and environment are listed in Sections 6, 7 and 8. Several of these postulated events required detailed modifica-tions to the present procedures, equipment, and structures as given in Section 9. These proposed modilications were based on a multilevel concept of system safety: (1) modifications of the crane coupled with detailed handling procedures help eliminate the possibility of a cask drop; (2) limit switches and specific cask movement paths minimize possible endangered areas and re-strict the maximum drop heights; (3) potential impact areas in-side these restricted areas are protected if necessary and the fuel bundles relocated to provide a further safety factor. J" 2-1

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3.0 DESCRIPTION

OF CASKS, CRANES, HANDLING PROCEDURES, STRUCTURES AND ECUIPMENT 3'l Casks (S) Spent Fuel Shipping Cask The spent fuel shipping cask is a cylinder 5'-2k" in diameter and 10'-84" high with a 2" thick stainless steel outer body and cooling fins. Cask diameter is 5'-10k" including the fins. The weight of the loaded cask is 120,000 lbs. It is raised by a single hook lifting yoke weighing about 700 lbs. (F) Fuel Transfer Cask (Refueling Transfer Cask) The fuel transfer cask lower portion is a cylinder 3'-0" in diameter and 8'-2" high. The upper portion is a cylinder 2'-0" in diameter and 5 '-8 " high. The total height of the cask is 14'-8". The weight of the loaded cask is 48,000 lbs. (C) Cobalt Cask The cobalt cask radioactive shipping container is a cy-linder 3'-1-3/4" in diameter including the cooling fins, and 9'-5-3/4" high. The weight of the loaded cask is 30,000 lbs. (T) Fuel Rod (TREAT II) Cask The fuel rod cask is a cylinder 2'-0" in diameter and 9'-8" high. The w'eight of the loaded cask is 15,0C9 lbs. e 3-1

3.2 Scent Fuel Handling Cranes and Hoists ), 3.2.1 Containment Single Leg Gantry Crane (75 ton) a. This crane was rated as follows: Hoist Trolley Bridge 6.5 fpm - 5 steps 5 fpm - 3 steps 50 fpm - 3 steps 1.2 fpm - 1 step 1.2 fpm - 1 step 40 hp @ 1200 rpm 5 hp @ 1200 rpm 10 hp 6 900 rpm 1/6 hp @ 1200 rpm 1/3 hp 9 1200 rpm Number lifts per hour: 3 Lifts: 62'-0" Span: 36'-4" Bridge travel: 74'-0" b. The crane was designed and constructed in com- )A() pliance with the requirements of: American Institute of Electrical Engineers American Standards Association ~' American Welding Society Electric Overhead Crane Institute - Speci-fication 49 - Class I National Board of Fire Underwriters National Electric Manufacturers Association State of Michigan Regulation for Cranes and Hoists c. The following minimum factors of safety were used in the design: 1. Hooks, shear blocks, bridge and trolley drives, (/ complete hoisting mechanism, trolley frames s_ and structural steel parts, not including bridge girders........................... 5 2. Bridge girders in accordance with ECCI No. 49 3-2

3. Welds: Bridge girders............... 4 Other......................... 5 II 4. Rope: No t le s s th an................. 5 d. The crane is cab operated e. The crane hoist is double reeved with 12 parts of 1"-6x37 wire rope. f. The crane is fitted with safety slings of restraining cables and shock absorbers capa-ble of stopping and holding the 30 ton refuel-ing cask after a fall of 20 feet within a dis-tance of 3 ft. g. The hoist has a mechanical drag (Weston type) brake which also functions as a holding brake. p. ya h. The hoist has a magnetic shoe type holding brake. 1. The hoist has one load sensing eddy current type drag brake. j. The trolley has one holding brake. k. The bridge has one holding brake and one fect-operated hydraulic brake. 1. The hoist motor has a thermai overload relay and internal thermal overload protection on two phases. m. m. The hoist has a screw type limit switch which controls the upper and lower limits of the hook. A manually operated by-pass switch is provided for the lower position. e_ _ 12

. _.. ~..,. ...... - - ~... - 3.2.2 Spent Fuel Handling Loading Crane a. This crane is used to load the spent fuel ). cask for transporting off-site. b. The crane is rated the same as the single leg gantry crane as described in paragraph 3.2.1-a. c. The crane was designed and constructed in ( compliance with the same requirements as the single leg gantry crane as described in para-I graph 3.2.1-b. d. The trolley has the following features: 1. Pendant controlled. 2. Trolley has one speed. 3. Hoist has two speeds. )( ) 4. Hoist has mechanical drag (Weston type) brake which also functions as a holding brake. 5. A magnetic shoe-type holding brake. 6. Trolley has a holding brake. 7. Upper and lower screw type limit switch. e. This trolley and hoist were rated as follows: Hoist loading rating: 75 tons Speed / Motor rating: Ecist Trolley 8/4 fpm - 2 step 15 fpm - 1 step 50/25 hp @ 1200/600 rpm ik hp @ 900 rpm / Lift: 26'-10" 3-4 = - :== -

i 3.3 Handling Precedures The following information was excerpted from the Big Rock Point Cask Operating Procedures. 3.3.1 shippinc cask A plan view of the pertinent elevation is included as Fig. 5-1. The cask (loaded on a dolly) is brought in at elevation 599 '-5" through the equipment hatch which is then closed. After connection of the lifting yoke, the shipping cask is pivoted to a vertical position and slowly raised and moved laterally using the 75 ton gan-try crane into the decontamination room at elevation 621'-0". A check of the cask is followed by removal of the cask cover. The cask is then moved directly to the spent fuel pool, and then lowered to the resting area in the southwest corner of the pool. 7 ~S u(%J While the shipping cask is moving from the decontamination area to the spent fuel pool, it is kept within approximately 6" of the floor except when passing over an obstruction near the reactor vessel. At that time it is raised approximately 5 ft. above the floor. All fuel shipping occurs during reactor operation due to the limited availability of cranes during outages. 3.3.2 Transfer Cask The fuel transfer cask is stored inside the containment near the spent fuel storage pool at elevation 624'-9". If the transfer cask is to be used on or above the reac-tor vessel, a set of safety cables is used to secure the ( cask in addition to the lifting slings. The double rigged cask is leveled, checked and then moved to either the reac-tor or spent fuel pool using the 75 ton gantry crane. [ ., -... ~ - -.. -. ~

~ ~~ ~ , _. _ _ [_ _ _ m.1 _. _ _ ] .y_ _ _ j_ _.._. Figure 5-1 shows the path taken by the cask during fuel handling. )p. 4 s. To remove fuel from the core, the refueling platform is rotated to a position allowing a vertical lift. The transfer cask is lowered into the extension tank to a depth sufficient to cover the fuel when the fuel is hoisted inside the cask". The cask lower door is opened and a fuel grapple is lowered by cable into the reactor. An operator, using a long actuator pole, attaches the grapple to the fuel bundle to be removed. The fuel is hoisted out of the core guided by the operator. When clear of the core, the actu-ator pole is removed and the fuel is hoisted into the cask. The lower door is closed and sealed and the cask is hoisted and moved to the fuel storage pool. At the storage pool the cask is positioned over a storage rack and lowered into the water. The cask door is opened ){]) and the fuel lowered into t he rack, guided manually by an operator using an actuator pole. The fuel grapple is then removed from the fuel. New fuel may be inserted into the core during the return trip with the transfer cask, or the monorail crane may be used in moving unirradiated fuel directly from the new fuel storage area into the core. Monorail crane lowering speeds are restricted to be consistent with the speeds of the other refueling cranes. 3.3.3 Cobalt Shipment Cask The cobalt shipment cask is used to transfer irradiated . cobalt from the spent fuel pool to an offsite area..The cask is brought in using the same dolly system utilized , () by the fuel shipping cask. The gantry crane is used to move the cask directly to the southeast corner of the spent fuel pool, passing north of the reactor. Figure 5-1 shows the approximate route of the cask. It is then moved to

the side of the fuel pool. The reactor is in operation while the cask is being moved across the refueling floor. I It is also moved approximately 6" above the floor. 3.3.4 TREAT II Cask The TREAT II cask is used to ship special fuel rods to Ge.neral Electric. This cask is brought into the contain-ment to be filled and uses the same dolly system utilized by the spent fuel shipment cask. The TREAT II cask is brought to the edge of the pool by the same route followed by the cobalt cask. Upon arrival at the edge of the spent fuel pool, the cask is prepared for loading. Decontamination procedures similar to those used on the cobalt cask are then followed and the cask is moved back to the dolly and outside of the containment. While the cask is being moved near elevation 632'-6", it is moved approximately 6" above O the floor. p l l 9 s_ i lN " ', - + - mged en a sh- - - = =,m, e - m er W em mm,

_.~... _.... _ 3.4 STRUCTURES g,, The general layout of the reactor building interior is shown in Fig. 3-1. The plan Fig. 5-1 and two sections, Figs. 5-2, and 5-3 of the reactor building show the location of the structures which may be impacted for the postulated cask drop cases. Table 3-1 shows dimensions and properties of the impacted structures..The concrete structures listed have a minimum strength (f') = 3,000 psi; an elastic modulus (E) = 3,000 ksi. The reinforcing has a specified minimum yield stress (f ) = y 40,000 psi. The following impacted structures were analyzed: (1) Spent Fuel Fool Floor: The 6'-0" thick rectangular floor-slab of the spent fuel storage pool is covered by two 1/2" thick lead sheets and 6" grout. The pool is lined with 3/16" stain-less steel. The slab is supported at all four edges, and is reinforced with ill bars 12" each way at the bottom and #7 bars 12" each way at the top. s ~. i (2) The fuel transfer cask storage rack platform at elevation 624'-9" is a steel frame consisting of c15x33.9 channels and 6x1/2" flatbars. It is anchored to, and cantilevered from, the south wall of the fuel pool. Material is ASTM A-7, nominal yield point (f,) = 33,000 psi. (3) The refueling floor at elevation 632'-6" south of the reactor, is a 5'-3" thick reinforced concrete slab which spans \\ from the reactor cavity shield wall to the wall north of the steam drum enclosure. The slab is reinforced by #9 bars 12" each way at the bottom. (4) The two concrete hatch covers in the reftaling slab south of the reactor are 5 '-3" thick, 5 ' 3"x6 '-0" at the base. Their structural properties are the same as for the refueling floor slab. 3-8

The reactor shield plug located directly above the top (5) head of the reactor vessel is 3'-6" thick, 15'-5" diameter reinforced concrete slab supported on the reactor cavity shield walls. A 12-1/2" thick, 16'-0" diameter lead lined reactor top shield is located on top of the shield plug. The top of this plug is l'-0-1/2" above the refueling floor slab. A 2 '-2" thick and 6 '-4" by 6 '-3" slab located 4 '-2-1/2" above the refueling floor, shields the 4" redundant core spray line south of the reactor. The reactor shield plug is reinforced by #9 bars 12" each way at the bottom and #6 bars 10" each way at the top. The top each face. shield slab has #6 bars at 12" each way at The cask decontamination area concrete slab is a 2'-6" (6) thick rectangular reinforced concrete slab 9'-0" wide by 13'-3" long, supported and restrained on all four edges at ) ~/ elevation 621'-0". Reinforcing is #9 bars 10" at the bottom and #9 bars 8" at the top. (7) Floor slab at elevation 599 '-5". The'laydown area ficor slab is a 3'-9" thick reinforced co'ncrete slab spanning 14'-9" between walls. The slab is reinforced by #9 bars spaced 12" on center each way at the bottom. (8) A steel hatch cover 4'-6" by 6'-6" is located in the lay down floor, top elevation 599'-5" close to the present path of the spent fuel shipping cask. The hatch cover consists of Material WT7"x15" T-beams welded to the 1/2" cover plate. conforms to ASTM A-7; nominal yield stress is f = 33,000 psi. 7 (9) The spent fuel storage racks are fabricated from light (- (3/16" and 1/4" thick) aluminum (6061-T6) framing members. j Two types of racks are used: s- ~ T

Spent Fuel Storage Racks Type A and B: 1 These racks consist of 3"x3"x3/16"x7' long vertical angles forming the corners of 7" square cells spaced one foot on centers. These cells are laterally stayed top and bottom by 3"x2"x1/4" horizontal angles and interconnecting 3"x3/16" bars welded to the vertical angles. Each cell has a 7-1/2" square 5/16" base plate. The rack is supported (on the bottom of the fuel pool) by 1/2" PVC pads under every other base plate. Spent Fuel Storage Rack Type C: This rack consists of vertical 6"x6"x3/16" fabricated struts, 5'-11" long, spaced 14-1/2" on center in an offset pattern interconnected by 3"x3/16" bars top and bottom. Perimeter vertical struts are 3"x2.33 lb. structural tees. Corner struts are 3"x3"x3/16" angles. The 3" bars and vertical struts are connected to a 3"x2"xl/4" angle extending around the perimeter on top, and a continuous 3/16" plate at the bottcm. The rack is supported on the bottcm by a 1/2" PVC sheet bearing on the floor of the fuel pool. Details of these racks are shown in Figures 3-2 and 3-3. (10) Fuel Cask Loading Platform: The fuel cask loading plat-form and supporting structures are located outside the reactor building. These structures were not analyzed because there are no safety related equipment at these locations. O / T .i _ _ 3-10

~ 3.5 Safety Related Ecuio=ent h The cooling water pumps, spent fuel pit pumps, control rod drive pumps, fuel pit heat exchangers, cooling water heat exchangers, core spray tank and a part of the core spray line are located below the possible movement paths of the casks. A short description of each of these items is included. The equipment location is shown in Figs. 3-4 and 3-5. 3.5.1 Coolina Water Pumos The cooling water pumps are the main components of the cooling water system. The reactor cooling water system is a closed intermediate cooling loop utilizing deminer-alized water to remove heat from the following pieces of equipment: )- Reactor Shield Cooling Panels Reactor Cleanup Non-Regenerative Heat Exchanger j Reactor Shutdown Heat Exchanger Fuel Pit Cooling Water Heat Exchanger Miscellaneous Sample Coolers Reactor Recirculating Pump Coolers The pumps are vertical, centrifugal type, each having a 1500 gpm capacity with a differential head of 100 ft. 3.5.2 Spent Fuel Pit Pumos The spent fuel pit heat exchangers are provided with a total of 500 gpm from two horizontal centrifugal pumps. The pumps may be operated individually or in parallel. / ' / ~

~... _.. 3.5.3 control Rod Drive Pumes )i There are 32 bottom-entry, hydraulically-operated control rod drive mechanisms spaced on 10.446-inch centers, each f drive actuating an individual control rod. Reactor feed-water is used as the working fluid for both normal and scram operations. Scram or emergency insertion of the rod is accomplished by similar drive flow paths as normal raising of the rod, except that the rate of control rod insertion is con-siderably faster. Thirty-two gas-water accumulators, one for each control rod drive mechanism, are the sources of the hydraulic pressure required for the scram at lower reactor pressures, while a shuttle valve within each drive mechanism admits reactor water to the drive when the reactor pressure ex-J- ceeds accumulator pressure. The accumulators are charged with nitrogen gas to a pressure which will deliver enough water to scram a fully withdrawn rod at low reactor pres-sures. The charge en each accumulator is monitored con-tinuously by a pressure switch. The water discharged from the drives during a scram is collected in a scram dump tank which is initially at atmospheric pressure. The accu-mulators and dump tank are isolated from the normal driving circuit connected to each drive by scram valves which are held closed by solenoid-operated pilot valves. d Scram is initiated by de-energizing the. scram pilot valves which allows the scram valves to open, thereby connecting the drives to the accumulators,and scram dump tank. The large differential pressure between the accumulators and dump tank rapidly drives the rods into the core. 3-12

3.5.4 Funi Pit Heat Exchancars ) There are two 125,000 lb/hr heat exchangers with a com-6 btu /hr. Individually, they are ca-bined duty of 6x10 pable of maintaining the water temperature of the spent fuel poel below 120*F while removing the decay heat from one-half of a fully irradiated core. For greater heating loads (more than one-half of a fully irradiated core), both heat exchangers will be operated. The heat removed from the spent fuel pit is then transferred to the cooling water system. 3.5.5 Cooling-Water Heat Exchangers The cooling water heat exchangers transfer the heat from the cooling water system to the service water system. They are an integral part of the shut down cooling system. g,. The service water system is an open system in which strain- "(. ed water is supplied from pumps in the intake structure and returned to the lake along with the discharge from the cir-culating water system. 3.5.6 core Spray Tank The core spray tank is provided in the core spray re-circulation system so that the pump suction conditions and the flow characteristics of the system can be periodic-ally tested. The tank also serves as a chemical add tank for mixing rust inhibitors into the lower portion.of the suction line. The core spray cooling system is provided to prevent a core meltdown should the core become uncovered following an incident. Initially the system is supplied from the )- fire protection system, followed by recirculation through l the core spray pumps and heat exchangers. Cooling water i is admitted into a circular sparger above the core which i directs sprays onto the fuel elements. /

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3.5.7 Core Spray Line ) -- The core spray ring and core spray nozzle inside the reactor vessel are supplied by two lines which enter the vessel from its southern side. These lines are shielded by protective concrete blocks. 3.6 Eauipment Tests A summary of equipment tests is included as Appendix B. f A .^.- i i l s 4 O e 3-14 _. ~.. _.... .y-- rr -y. w w---

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V l IJ ,i t } e s O s l 0 ~ .s / 7 TABLE 3-1 STRUC111RAL DESCRIPTION / / t Steel Concrete Specified Hodulue n T$d Top SPAN THICKNESS Concrete Yield Of Resistance Displacement UU* Elev. Length Width Strength Stress Elasticity Reinforcing At Yield At Yield Itazimum E3I Location (ft) (ft) (ft) (ft) (kst) (kst) (kst) Steel (k) (in) Ductility mud 1 S.V. corner of spent 602 '1" 26.0 20.0 6.00 3.0 40.0 3000

11012" E.W. Bot,

8500 0.010 30 I

7012" E.W.

Top fuel pool slab Center of spent fuel 602'-1" 26.0 20.0 6.00 3.0 40.0 3000

11012" E.W. Bot.

3560 0.025 30

7012" E.W.

Top pool slab 20 ' 2 Transfer cask, 624'-0" 3.0 canti-h" covert 33.0 230 0.37 storage platform lever 2-C15 x 33.9 3 Refueling floor 632'-6" 8.0 5.25 3.0 40.0 3000

9012" E.W. Bot.

5600 0.027 30 4; Hatch cover in 632'-6" 6.D 5.25 5.25 3.0 40.0 3000

6@l2" E.U.'.F.

refueling floor 5* Reactor shield plug 633'-6

15.41 (dta.)

1 50 3.0 40.0 3000

9012" E.W. Bot.

1050 0.048 30

6010" E.W. Top Top shield 633'-6 "

16.0 (dia.) 1.04 3.0 40.0 3000

6012" E.W.E.F.

1050 0.048 30 6 Decontamination 621'-0" 13.25 9.0 2.50 3.0 40.0 3000

9d8" Top 2900 0.054 30 area of floor f9010" Bot.

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9012" E.W. Bot.

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{'

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See Amendment #10 (Ref. 4)

See Table 6-5 for Impact Case 9: Cank prop on Fuel Rocks

4.0 SPENT FUEL HANDLING CRANES AND HOIST INVESTIGATION [ 4.1 Structural 4.1.1 Structural Analysis The single leg gantry crane and the spent fuel cask loading hoist trolley were structurally designed to have the following structural factors of safety: Bridge girders - EOCI No. 49..................... 4 Hooks, blocks, trolley frames.................... 5 Welds - Bridge girders........................... 4 Other............................................ 5 4.1.2 Loadine The equipment was designed for the following conditions: Gantry crane and fuel cask loading hoist trolley: Rated load - 75 tons Impact (15%) - 11.75 tons ~ a i EOCI No. 49 - Class I 4.2 Mechanical 4.2.1 Stress Analysis The mechanical components were designed so that the stress limits are within the design allowables.when subject to the loads given in Section 4.1.2. The analysis is included in the structural calculations. The conditions are the same as for the structural analysis. p(-- 4-1

b. The following factors of safety were used in the de-sign of the mechanical components: (

Load carrying components...................... 5
Welds.........................................

5

Rope: Not less than........................... 5 4.2.2 Brakes The gantry hoist has two holding brakes; (one mechanical a.

type and one electrical type) and one drag brake. The mechanical (Weston type) holding brake serves as a back-up drag brake. The drag brake is of the eddy current type. b. The spent fuel cask loading hoist trolley has two holding brakes; one electrical type and one mechanical g. type. The mechanical (Weston type) brake serves as the drag brake. c. Each trolley has one electrical holding brake. i d. The gantry bridge has an electro-hydraulic combined holding and drag brake. e. The electrical shoe type holding brakes are rated to j hold the full load suspended when power to the hoist motor is off. f. The holding brakes for the 75 ton hoist are Westing-house magnetic type SA-direct current, frame size SA-1355, style number 1819-221 rated at 400 foot i D .E ~ l 4-2 r v r

,-~;___.______.._ f. (Contd) pounds torque. These brakes have a direct current )~ clapper type magnet and are designed so that when the magnet is energized the shoes will clear the wheel and when de-energized the shoes are pressed against the wheel by means of the compression springs. They are self-adjusting and the worn lining torque is not loss than 10% of new lining torque. The torque rating is marked ou an indicator and may be adjusted. l 9-The holding brakes for the 75 ton hoist are mounted on the gear drive shaft. The hoist motor is rated at 40 hp,1200 rpm, with five step control in either direction. The rated full load torque for his motor is 175 f t -lbs. The normal pull-out torque a wound rotor motor is 225 to 275% of the full load rque. The motor stall torque would be 482 ft-lbs maximum which exceeds the brake capacity of 390 ft-lbs minimum. 5# h. The electrical hoist holding brake will not prevent two blocking of the hoist with no load; however, with rated load on the hook and the brake set the motor will stall and the overload protection devices will disconnect the electrical power to the motor. There are two overload devices; one is an overload relay, the other thermal overload switches, in two phases, within the motor. i. In the event of the failure.of the electrical holding brake for the hoist to set and with continued operation of the motor with full ra'ted load on the hook, the hoist will two-block. The load on the rope at the equalizer sheaves would be approximately 50,000 lbs. This results in a factor of safety of 1.6 based upon the rope strength. This load on the rope does not in-clude stress increase at the equalizer sheaves. 4-3 -. ~._.

.~ ~ 4.2.3 Two Blocking I-a. Either of the following conditions could result in two 7 blocking of the 75 ton hoist:

1. Failure of the upper limit switch while hoist con-trol is in hoist position.

2. Continued operation of the motor, in the hcisting direction, with hoist control in neutral position.

b. The factor of safety of the rope at the equalizer sheaves will be approximately 1.6 for either of the above conditions with rated load on the book. c. With no load on the hook the factor of safety for the rope at the equalizer sheaves will be 3 for either of the above conditions. s / 4.2.4 Hoist Drive System a. For the 75 ton hoist, the hoist drum is driven by a 62:1 ratio gear box which contains the mechanical brake. The electrical holding brake is attached to the input shaft of the gear box. The motor is connected to the electric holding brake by a cross shaft with two coup-lings. The eddy current drag brake is attached to the i back of the hoist motor through a shaft coupling. b. The hoist drum is mounted on pedestal bearings supported on each trolley truck assembly. The upper sheave assembly is supported from cross girts c. between the trolley trucks. v ).s i d. The hoist drive motor and gear box is attached to the trolley deck. ae w ., +,.

f ~T m .s Commission, Safety Standards, Part 18, Overhead and Gantry Cranes. Tests and inspections are scheduled ),m for equipment in continuous and standby service. III. SPECIAI. TESTS AND INSPECTIONS A. The crane hooks and clevis pins were satisfactorily exa$ined by the volumetric ultrasonic examination procedure during March 1972. The 75 ton hook and the 5 ton hook were examined by the liquid penetrant tech-nique and the magnetic. particle technique with no indications on March 22, l'974. B. Visual examination of a fuel transfer cask was per-formed for a possible weld crack in one of the ears on the front side of the transfer cask. Visual exa-mination revealed that a liquid (paint, lubricant, etc.), black in color, had run down the edge of the '( weldment accentuating a crack-like appeara. e along J the edge of the weldment's heat-affected zone. Fur-ther analysis concluded that the weldmentswas not cracked. j -J' B-3

4.3 Controls and Electr'ical ~ 4.3.1 Limit Switches y 4 Each 75 ton trolley hoist has a screw-type limit a. switch that sets the electrical holding brake at upper and lower hook travel limits. b. A system of trolley and gantry travel limit switches has been provided. These switches are used to set the holding brake at the ends of the trolley travel. The bridge brake does not set upon operation of bridge travel limit switches. 4.3.2 Hoist Controls The 75 ton hoists have five steps of speed in the hoist-a. ing and lowering directions. These controls are pro-vided with resistors for proportioned speed control. ,{~x The lowering speed at full hook load without the eddy current brake would approximate 120% of the hoisting speed in the same step. 4.3.3 Brake Control The eddy current brake for the 75 ton hoists are capable a. of providing an almos't infinite control of the braking 1

torque, i.e.,

lowering speed. The hoist controller in the cab has a slow-fast selection switch.ctached to the hoist control switch. The eddy cncrent brake is a Louis Allis Model AB706 which requires power input to operate. b. The electrical power for operating the holding brakes for the 75 ton hoists comes from the terminals of the ig_) hoist motor. The brakes are activated when power is off to the motor. l-l~ 4-5 w --i--------v-~. .--a. .--m - .,y

4.3.4 Trolley and Bridae Control s <~ a. The 75 ton hoist trolleys travel is controlled by a l three step controller. A slow speed drive is available through the use of a 1/6 hp motor, gear box and clutch. b. The gantry crane travel is controlled by a three step controller. A slow speed drive is available through the use of a 1/3 hp motor, gear box and cit cch. 4.3.5 Electrical a. The electrical systems furnished are three phase, 60 hertz, 440 volt power. b. The gantry crane has the following electrical switches 4 and protection devices: Disconnect switch (manual) ~ )(f Mainline circuit breaker Magnetic disconnect Each motor circuit is provided with: Overload relay Motor thermal overload cutout c. The 75 ton fuel cask loading trolley has the following electrical switches and protection devices: Disconnect switch (manual) Mainline circuit breaker Magnetic disconnect Motor circuit breaker Motor thermal overload cutout 4-6

d. Power supplies for controls are provided with circuit breakers. g s e. The power supply circuits are not grounded. .9 J 1 }'. 3 4-7

5.0 CASK DROP ACCIDENTS I The plan Fig. 5-1 and sections Figs. 5-2 and 5-3 show the location of the postulated cask drop accidents. Table 3-1 lists the impacted structures, their location, elevation, dimensions, properties, resistance, yield dis-placement, and maximdm ductility in accordance with Ref. 1. A brief summary of postulated drop accidents ar" given for each of the four casks. The accidents are identified by the case number assigned to the structure impacted. The impacted structures are numbered and described in para-graph 3.4. 5.1 Spent Fuel Shipping Cask Drop Accidents The spent fuel shipping cask has the following postulated h drop accident locationc: r l 5.1.1 Southwest Corner Of Spent Fuel Pool Case (1) - The spent fuel shipping cask may drop from one foot above the refueling floor slab to the bottom of the pool. 5.1.2 Fuel Transfer Cask Storage Platform Case (2) - Not applicable. The fuel shipping cask is 1 not routed over the storage platform. i 5.1.3 Refueling Floor Slab Case (3) - The spent fuel shipping cask may drop from i five feet above the refueling floor slab south of the 9-reactor to the refueling floor. The floor slab is lo-cated above the steam drum enclosure. l l 5-1

~ 5.1.4 Concrete Hatch Covers Case (4) - The spent fuel shipping cask may drop from g five feet above the refueling floor slab onto the con-crete hatch covers in the south side of the slab above the steam drum enclosure. 5.1.5 . Reactor Too shield Case (5) - The spent fuel shipping cask may drop from five feet above the refueling floor slab onto the edge of the shield which covers the 4 inch redundant core spray line and subsecuently overturn and fall on the reactor top shield. 5.1.6 Decontamination Area Floor Slab Case (6) - The spent fuel shipping cask may drop from one foot above the refueling floor to the decontamination I area floor slab at elevation 621'-0". 5.1.7 Flocr Slab At Elevation 599'-5" Case (7) - The spent fuel shipping cask may drop from one foot above the decontamination area floor slab onto the floor at elevation 599'-5" near the fuel cask dolly track. 5.1.8 Steel Hatch Cover Case (8) - The spent fuel shipping cask may drop from one foot above the decontamination area floor slab onto the steel hatch cover located in the floor at elevation 599'-5" south of the fuel cask dolly track. \\ 5-2

5.1.9 Fuql Storaca Racks case (9) - The spent fuel shipping cask may drop from one foot above the refueling floor at the spent fuel pool, hit the edge of the pool and fall onto the fuel storage racks type (B) located in the west side of the pool. 5.2 Fuel Tr'ansfer Cask Drop Accidents The fuel transfer cask has the following postulated drop accident locations: 5.2.1 Floor Of Spent Fuel Pool Case (1) - Same drop location as described under section 5.1.1. 5.2.2 Fuel Transfer Storace Platform g.. J Case (2) - The fuel transfer cask may drop from one foot above the refueling floor to the cask storage platform south of the pool. 5.2.3 Reactor Core The fuel transfer cask may drop from above Case (5) the refueling floor and into the reactor core. An ana-lysis of this drop accident has been performed. (Reference Amendment No 10 to the Final Hazards Summary Report.) 5.2.4 Fuel Storace Racks Case (9) - The fuel transfer cask may drop from one foot above the refueling floor to the fuel storage racks type ,[ (B) in the west side of the spent fuel pool. g 1 5-3 -r

5.3 Cobalt Cask Drop Accidents ).- The cobalt cask has the following postulated drop locations: 5.3.1 Floor Slab At Elevation 599'-5" Case (7) - Same as described under Section 5.1.7 above. 5.3.2 Fuel Storace Racks Case (9) - The cobalt cask may drop frem an elevation of one foot above the refueling floor slab to the fuel storage racks type (A), (B) or (C) located in the north side of the spent fuel storage pool. 5.4 Fuel Rod (TREAT II) Cask Droc Accidents \\ The same drop locations are postulated for the fuel rod cask as for the cobalt cask listed above. See plan of postulated cask drop locations, Fig. 5-1. h \\ J '-) 5-k

i' N 5 SPENT FUEL SHIPPING CASK F X FUEL TRANSFER CASK C -.- CO B A LT C A S K l ) T -- FUEL RCD (TREAT ~:") CASi [ h...(hPCSnJLATED CRCP ACCIDENTS'! l .- Ecut PM ENT g lll AIRLOCK / I LAYCCWN AREA ~ \\ '/ I CECCNTA UNATICd l AREA FLbCR j N. 4 EL.62 t'- 0" i ~ Il ll, h FUEL DCLLY TRACK l STEAM CRUM ENC'JSURE', i QM N-a REFUELING FLCCR E=.--- a c.,, l REACTCR 4 fl s N TCP SHIELD ' ($; / L5] \\k.. . i [ =- d _- - _a ~~ E I 1, l

q

q i X -'I L;_.. I. g , --{ X fl ~Y " f(9 I - RACH B 1 a r 'll, i .s ,.~ m --RACKS A S B g I/ \\ 7 o$ mD ' l. l f l ~ RA CK C y n N h, o -l FUEL TRANSFER CASK STORAGE i -SPENT FUEL PCCL PL ATFCRM EL. 624'. 9" /'/a FLOC R EL. 60 2*- t " U ,\\ /

i_

/ u u- -.-.--.-- a FIGURE 5-1 Plan of Postulated Cask Drop Accidents Existing Routing Of Casks Indicated

~ S SPENT FUEL SHIPPtN3 CASK / REACTOR BUILDING F,, X FUEL TRANSFER CASX 0 ---COSALT CASX T ~~ -F1JEL RCD CASK W! 'l i f a SERVICE GANTRY CRANE h' I i i 4 - r J im 4 i ~~ ! W, ',, F' jij k ~ i is: J 75 TCri CRANE 5 Tott \\ MONORA!L CRANE s -%Q \\ T-l l,d l'r ~ f.g lt (REFUELING ,FLOCR EL.632 L i _ //- x_ r / - i li-IE STORAGE oun: ev e 1c;nt-s' ' J l p[-SPENT FUEL l k-h)I N f ED '.7. 1*

c J

J./ /// )( l t i i e N u f. : K l . n, / ' 1. ; - RACKS ,j 't 9 \\ @l ,9 : mm l 1 ~ EE l -l 3, i Y EL 600'- 6" i / \\ / %p.' g a s ..f ' y l- '] \\ y' @ h ( @ CD [ EL. 5 85 '- 6" +, .e.. s. ' 3:. y...- ,'. 6 , q. : - l FIGURE 5-2 North-South Section Through Reactor Building And Spent Fuel Pool. ) Postulated Cask Drop Accidents 1, 2, And 9 Indicated. l i

) s [REACTORBLDG. S S S. 8 I l.^ REFUELING F1.00R EL.632'-o ^ 'r i i i..,...i 4 s. ,r Y^' -l 'E ****.T~ 3-l > ' o' # DECONTAMINAT M i.' AREA i l l l i t -G-q, W e,.. e,, i i l j .s g 1 l l q e [7' T I I SPENT FUEL "~ M; gt;rm,- / ll 0

.3.

M E ~f Ir ...td G l. I pool l n. .g [ -f ,p d l l i ul! 'I l , o s, d t ~ g t I a u ; i.: ( . v.. {i l ? I G l ii l EL.602'- l" r-- ! f- .EL.6CO'-6? iY l b' 3i M ~~ ~ i ~' 3 ~" , ;,, J( g a l C a ,g s q .,s ~.. i s ~.i.. r-3-- /-,

N FUEL. PM H.E.8 hl EL. SQS _9,%_ ___,j _l l

l ~ $9 ) l J. EL. "9 S'- E ) --t r - _ _. _l e : 1..-lI..,:,- 1: I CONTROL RCD,CR'iVE l I pd l PUMPS 4 i 1,g',- I i i l i . q.k :- l l,. = EL. 575,- 6,, l ^ t 1 (;,.. t f U f

y 1 1

I ly:; ~. ". ', g... g,.,..,. 1 FIGURE 5-3 East-West Section Tbrough Reactor Buildinc }'- Postulated Cask Drop Accidents 3, 4, and 6 9 Indicated. 3 nD D M\\ \\D

~. ~ 6.0 EFFECTS ON STRUCTURAL INTEGRITY }~ The methods and criteria described in Bechtel Topical Report BC-TOP-9 (Ref. 1) were used in this paragraph to predict struc-tural failure. Tables 6-1, 6-2, 6-3, 6-4, and 6-5 list the drop accidents postulated for the four casks. Drop heights, impact velocities, energies, strain energy capacities, and failure criteria (based on Ref. 1) for each of the structures impacted, are listed in the tables. The tables show that all postulated cask drop accidents, sxcept 3 and 4, will cause structural response which exceeds the cri-teria for acceptable behavior. The energy absorbing capacity of the structure will be exceeded as determined by calculations performed according to BC-TOP-9 methods (Ref. 1). The drop accidents are described in paragraph (5). ? 3 The calculated effects upon the impacted structures are listed in the following: Case (1) - The spent fuel shipping cask drop on the floor slab of the spent fuel pool - The floor slab will crack in bending and the 3/16" thick stainless steel lining may be perforated. The crack size could cause pool leak exceeding the 200 gallons / minute capacity of the pool make-up systems. Slab penetration is not considered possible. Case (2) - Transfer cask storage platform - The postulated transfer cask accident will collapse the storage platform, the members would yield and/or the anchors pull out of the wall. ~ Cask and platform would then drop and possibly tumble onto the floor below at elevation 600 '-6" with sufficien't residual energy to collapse this floor and impact the floor below at elevation 585'-6". 6-1 ~

'h* ^ ' + e Cases (3) and (4) are not critical provided the shipping cask drop height is limited to less than or equal to 5'-0". i Case (5) - The reactor top shield may crack and fail in bend-ing subsequent to the postulated impact of the spent fuel shipping cask (paragraph 5.1.5). 1 The postulated drop of the fuel transfer cask into the core of the reactor vessel has been investigated earlier (refer to Amendment No. 10). Case (6) - Decontamination area above room 444 (floor at elevation 621'-0") - This floor slab will crack and possibly fail in bending. The residual energy could be sufficient to cause a chain reaction of failures: a) Failure and collapse of ficar slab at elevatien 621'-0". )(~ b) Failure and collapse of floor slab at elevation 599'-5". ~ Case (7) - Ploor at elevation 599'-5" near the cask dolly - This floor will cellapse, causing the fuel removal dolly track bridge and the dolly to fall with it, thereby providing enough energy to collapse the floor below at elevation $85'-6". Case (8) - The steel hatch cover will buckle and collapse together with the floor. Similar effects as described under l l Case (7) may be expected. Case (9) - Spent fuel storage racks - For the routing shown in Fig. 5-1, three casks (fuel transfer, cobalt and fuel red (TREAT II) can be moved over the fuel racks. r' 1 <J 5-2

The cobalt and The fuel transfer cask is routed over Rack B. I fuel rod (TREAT II) casks are routed over Racks A, B and C. For this analysis, it is assumed that the cask drop starts one foot above ficor level (three feet above water level). The casks would fall a distance of 22 feet through water before contacting Racks A and B and 22 feet before contacting Rack C. The kinetic energy at contact for each cask is shown in Table 6-5. The strain energy capacity of the impacted cells and the number of cells engaged are also tabulated. The strain energy capacity of the racks is limited by elastic buckling of vertical elements loaded in axial compression by cask L pact. The values shown in Table 6-5 are upper limit estimates based on the Euler elastic buckling formulae. The casks were assumed to fall on the interior portion of the Y^ rack (away from edge members) engaging a minimum number of J elements. Racks A and B Racks A and B are subject to impact by three casks (fuel trans-fer, cobelt, and fuel rod (TREAT II]). The casks would first engage the cells with the 1/2 inch PVC pads between the base plate and the pool floor loading the vertical members in axial compression. These cells have insufficient strain energy ca-pacity to stop any of the casks (for the drop heights' considered) and would fail in an elastic buckling mode. Subsequent to the failure of the cells with PVC pads, a cask would proceed down-ward forcing the cells without pads against the fuel pool floor. The vertical elements would be loaded in axial ccmpression and fail in an elastic buckling mode. These cells.also have in-sufficient strain energy to stop any of the casks from the pos-e ) tulated drop heights. ~ 6-3

Rack C Rack C is subject to impact by the cobalt cask and the fuel rod (TREAT II) cask. The vertical elements would be loaded in axial ecmpression and would fail in an elastic buckling mode similar to the cells in Racks A $nd B with PVC pads. The strain energy required to produce buckling of the engaged elements is considerably less than the kinetic energy of the casks at impact. Racks A, B, and C Subsequent to buckling of the vertical support elements, the dropped casks would continue relatively unretarded toward the bottom of the pool and likely fail adjacent elements (connect-ed by upper lateral restraint members) by dragging them la-terally and downward. ) The additional energy absorption by this failure mode would be minimal and would be further limited by possible failure of connections of members undergoing large' deflections. Considering the foregoing, it is concluded that if a fuel rack is subjected to any one of the postulated cask drop accidents, support for the spent fuel elements would be lost.

Further, the cask (s) would proceed to the bottom of the pool with only slightly diminished kinetic energy.

Similar damage would be expected should a cask first strike the edge of the spent fuel pool and then be deflected onto the spent fuel racks. The possibility of the fourth cask (the spent fuel shipping cask) tumbling and falling onto fuel storage rack (B) sub-sequent to hitting the edge of the fuel pool was considered r and evaluated. The consequences would be similar to, or..less severe than, those described for a straight, vertical drop ~ of the fuel transfer cask. 6-4

The previous predictions of structural failure were based )( on the criteria given in Ref. 1. It is possible that'for cases where the energy ratio is greater than one, but less than four or five, failure may not result because more capacity and deformation capability exists than assumed in the calculations. In addition, more effective mass may be available than assumed. e 4 t. I O P i j 6-5 _1

Ys V i f TASLE 6-1 SPENT FUEL SHIPPINC CASK (See Tige. 5-1, 5-2, and 5-3 POSTULATED CASK DROP ACCIDENTS for impact location.) Irpacg Energy Input Structure Ratio Of input Lccation Cask Weight _ Cask Droo Heirht in. Velocity Energy Transmitted Available Energy To Available Ccsa Including Yoke Air Water ToIgl____ At Impact At' Impact To Structure Strain Energy Capacity Energy St ructure Ms. (k) (ft) (ft) (ft) (ft/sec) (k-ft) (KE) (k-ft) (TE) (k-ft) KE/TE Failure 1 120.7 3.0 28.4 31.4 40.0 3050 1400 220 6.35*, Yes 3 120.7 5.0 0.0 5.0 18.0 600 350 37o 0.95 No 4 120.7 5.0 0.0 5.0 18.0 600 350 < 0.95 No 37o 2.60 Yes 322 124 5* 120.7 4.0 0.0 4.0 16.0 480 6 120.7 14.5 0.0 12.5 28.2 1490 1250 385 3.22 Yes 7 120.7 22.6 0.0 22.6 38.0 2720 2815 11 165 Yes 8 120.7 22.6 0.0 22.6 38.0 2720 2700 62 45 Yes

Reactor top shield plugs and head in place.

TABLE 6-2 FUEL TRANSFER CASK (See Tige. 5-1, 5-2, 5-3 g POSTULATED CASK DROP ACCIDENTS for location.) Structure Energy Input Available Ratio Of Input Irpact Location Cask Drop Height fa_ Velocity Energy Transmitted Strain Energy Energy To Available Structure Case Cask Weight _ Air Water Total At Impact At Impact To Structure Capacity Energy No. (k) Tit) (ft) (rt) (ft/sec) (k-ft) (xE) (k-ft) (TE) (k-ft) xE/TE railure M 1 48.0 3.0 28.4 31.4 40.4 1241 488 220 2.52 Yes 2 48.0 8.75 0.0 8.75 23.7 396 390 164 2.38 Yes SE A 5* 48.0

Reactor top shield plugs and head removed, see Amendment #10 (Ref.4)

I _g7,3 P

i k s i i i i TABLE 6-3 COBALT CASK (See Tige. 5-1, 5-2, 5-3 I POSTULATED CASK DROP ACCIDENTS for location.) i 4 Structure Ratio of Input Impact Energy Input Available Energy To Available Location cask Drop Hefeht in Velocity Energy Transmitted Strain 1:nergy Energy Structure Case Cask Weight Air Wter Total At Impact At Impact To Structure Capacity (TE) EE/TE Failure No. (k) llt) (ft) (it) (ft/sec) (k-ft) (k-ft) (k-ft) 7 30.0 34.1 0.0 34.1 46.8 1020 650 136 4.75 Yes t TABLE 6-4 IVEL R0D (TREAT II) CASK (See Figa. 5-1, 5-2, and 5-3 POSTULATED CASK DROP ACCIDENT for location.) h Impact Energy Input Structure Ratio of Input location Cask Weight Cask Drop Helpht in Velocity Energy Transmitted Available Energy to Available r, Case (k) Air Wter Total At Impact At Impact To Structure Strain Energy capacity Energy Structure No. (ft) (ft) (ft) (ft/ecc) (k-ft) (KE) (k-ft) (TE) (k-ft) KE/TE Failure Leio 7 15.0 34.1 0 34.1 46.8 510 s 300 136 2.2 Yee w c== h

w 'I / T .s -\\. l 4-s <f TABLE 6 5 CASK DROFS ON FU RACKS Kinetic ** Cask Cask No. of Velocities Energy at Kinette Fnergy Weight Diameter Uatts Drop Itcip.ht (fr.)_ at Impact Impact Strain Energy Case Rack ( (kips) (in.) Engaged Air Water Total (fps) (It-kips) Capacity (ft-kipe) Strain Energy Failure 9 A&B Tuei. 15 16 1 w/ pads ** 3 22 25 35.7 298 3.0 99.5 Yes (YREAT 41) 2 wo/ pads Cobalt 30 37.75 4 w/ pads 3 22 25 34.9 568 9.4 60.5 Yee 5 wo/ pads 9 C Tuel Rod 15 16 1 3 23 26 36.7 307 1.3 236 Yes (TREAT II) Cobalt 30 37.75 9 3 23 26 35.4 582 11.7 49.7 Yes 9 s Fuel 48 36 4 w/ peds 3 22 25 36.8 1013 9.4 108 Yes Transfer 5 wo/pada

Since the muss at the struck units is small compared to the mass of each cask, nearly all kinetic energy would be transiscred to ths, units.
- Every other cell of A and B racks is supported by I PVC pade.

s l

7.0 PISPONSE TO THE LOSS OF SAFETY-P2 LATED EQUIPMENT ) Table 7-1 correlates the safety-related equipment described in Section 3.5 with the postulated cask drop accidents spe-cified in Section 5.0. The response of the reactor system following the individual equipment loss is listed below. The 1 css of 'the spent fuel pool pumps and/or the spent fuel pool heat exchangers will not affect the operation or safe shutdown of the reactor. The fuel pool will be cooled frem two auxiliary systems supplying a total of 200 gpm of water which is sufficient to remove the decay heat of the fuel rods.if the pool is still intact.' section 8 examines the results follcwing violation of the pool integrity. The loss of the cooling water pumps and/or the cooling water heat exchangers would not substantially affect the operation of the reactor. The loss of the reactor shield cooling system )o. wculd require reascnably prompt shutdown of the reactor. The reactor may be brought down to a partial power level follow-ing the normal procedures, but the loss of the cooling water system would eliminate the use of the shutdown heat exchanger during the latter parts of the shutdown sequence. The emer-gency condenser could be used to remove the heat energy rar-mally transferred by the cooling water system. Normal reactor operation and shutdown are not dependent on the integrity of the core spray system. This system is an emergency cooling system for use for an LOCA (loss of coolant accident). The core spray tank is not an integral part of the core spray system and thus its loss is unimportant. e pv 7-1

- m-w=*ew w w e-.~

~~ The reactor cannot be operated without the control rod drive }. pumps. A redundant system (described in Section 3.5.3) will be able to safely shutdown the reactor with the normal control rod drive system inoperable. Thus, the loss of this system will not affect the safe shutdown of the reactor. Loss of water from the previously described systems, coupled with a complete emptying of the spent fuel pool would result in flooding of the containment to a depth of approximately 6 ft. This would not endanger any other safety-related sys-tems. -n f,, ~ \\.. 4 4 )v 7-2

1 %.~ IMPACTED STRUCTUPES 8 O 4 m O M 4 u u 4 n 3 u N C ^ .C e O M 1 00 0 U th 3 M D M U q w N M W. I C C4 3*. G W M A O k C m 4 O w C w 'O O m W Q O to O M +4 m O u to .O k M M U U w P: W ~4 4 =4

M 01 G

.C C 3 k M U C to U3 4 0 O G ~4 C 3 2 U .C4 O k c En 3 4 N M O u o U m U .C U C M C U C C 3 U u O O C o a w u a u o a c. u u n o a q c. v3 t a:

=

c: C w tn SAFETI EQUIFMDiT M N M a D e~ co e

1. Spent Fuel Fool Pu=ps W

W I r

2. Cooling Water Pu=ps l

W W 3 W W

3. Spent Fuel Pool Heat Ex.

~

4. Cooling Water Heat Ex.

W W

5. Core Spray Tank X

X X

6. Core Spray Line Z

X X I.

7. Control Rod Drive Pu=ps W

W

8. Shut Dcwn Heat Ex.

W X W

9. Shut Down Pu=ps W

X W X = Cask i= pact =ight da: age safety equip =ent W = Water leak =ight damage safety equiptent TABLE 7-1 CORRESPONDENCE BETWEFN ,( POSTULATED ACCIDENTS AND PLANT JAFETY EQUIEiETI F

8.0 ENVIRONMENTAL EF'FECTS Following a hypothetical cask drop into the spent fuel pool, two different types of radioactive releases could occur. The cask could crush some fuel bundles and breach the fuel rod cladding. A portion of the volatile fission gas from the gap and plenum regions of each fuel rod could be released to the fuel pool. 'An investigation of this accident, based on,AEC Safety Guide 25 results in the following site boundary doses. TABLE 8-1 SITE BOUNDARY DOSE PER CRUSHED FUEL BUNDLE DOSE (rem) Actual Des _icn Limit (10 _CFR 100) ~ Whole Body 6.43x10 " } 25 Beta (skin) 8.03x10~3 Thyroid .274 300 I The spent fuel pool can hold 120 fuel bundles, but only 1/3 of a.. core (28).cculd have the exposure hypothesized in the previous analysis. The remaining bundles (if they are in the'fu61 pool) would contribute much less to the site boundary dose.

Thus, a complete crushing of all the possible fuel bundles available would not violate current dose limitations.

The preceeding study was based on the continued integrity of the fuel pool. The water would absorb some of the iodine and provide cooling for the removal of the decay heat. Noble gases are not retained in the water and are assumed to be immediately released to the containment. A second type of accident which would result in a radioactive re-lease may be postulated. l 8-1

A cask drop'could result in a leak in the fuel pool which might ) be larger than the makeup capability of the backup systems. In this case the decay heat being generated inside each fuel pin would eventually melt the fuel pin cladding. The accumulated fission gases would be released directly to the containment atmosphere resalting in the site boundary doses shown in Table 8.2. TABLE 8-2 SITE BOUNDARY DOSE PER VIOLATED FUEL BUNDLE (No Fuel Pool Water Available) DOSE (rem) Actual Desian Limit (10 CFR 100) ~ Whole Body .257x10 2 } 25 Beta (skin) 8.03x103 Thyroid 27.4 300 45 Thus, if there are more than ten bundles in the fuel pool, a cask drop accident which violated the pool integrity could result in unacceptable site boundary doses.. l I l d'd 8-2 ~ ' ~ ~

9.0 RECOMMENDED MODIFICATIONS Based on the investigation of the censequences of a postu-lated cask drop, the folicwing modifications to equipment and handling procedures are recommended. 9.1 Upgrading Equipment and Procedural Controls Available statistics on cranes designed to the standards typical of those used in nuclear plants show the probability of an accident to be very small. A modification of the present crane supplemented by movement of the spent fuel bundles to a more remote location in the spent fuel pool, would further re-duce the possibility of damaging the fuel. Improved cask handling procedures would eliminate the possibility of damage to safety-related plant equipment. ) 9.1.1 Crane Modification 9.1.1.1 Containment Single Leg Gantry Crane To bring this crane into compliance with present day technology, the following modifications should be made: Provide an upper limit switch (paddle type). Provide a holding brake, with or without gear train,. .-.j 7[ on undriven end of hoist drum. (' Provide supports for drum to prevent hoist drum drop p 1 y '3 in the event of a drum shaft failure. . <F v] Provide a hand operated switch in cab to set holding brakes. Provide overspeed switch set at 120% of rated hoist- ' ~ 3 ing speed to set holding brakes. p Provide thermal overload warning indication for hoist motor. 9-1

7 Provide manual release device for holding brakes. Brakes are not to be capable of being locked into the released position. Provide for drag brake (eddy current type) to have self-excited auxiliary alternator to excite the eddy current brake and allow safe lowering of the load in the event of simultaneous failure of the holding brakes. Provide limit switches to prevent movemen of casks beyond the edge of the impact pad. 9.2 Fuel Pool Modifications Figure 9-1 shows the present and proposed locations of the fuel racks in the spent #uel pool. The relocation of the fuel racks, coupled with the limited crane movement into the spent fuel pool, would greatly reduce the probability of a cask drop on the spent fuel. The TREAT II, cobalt, and shipping casks \\~ may all be loaded in the west' side of the pool. J Since procedures require that the safety cables be attached to the fuel transfer cask before it can be moved, it will be allowed to move over the fuel racks when transporting fuel be-tween the fuel pool and reactor vessel. However, the fuel transfer cask will take the same path as illustrated in Figure 9-2 and will only move over the fuel racks while in the water. A segmented, removable pad of energy-absorbing material approxi-mately 20'-0" long by 6'-8" wide by l'-10" thick (see Fig. 9-3) is proposed for the western end of the fuel pool. This pad would protect the bottom of the fuel poc1 against impact loads resulting from cask drops. This, coupled with the modified handling procedures, would ensure that pool integrity would oe maintained following a cask drop in the spent fuel pool. w' 9-2

9.3 Cask Routing Modification I For simplicity and safety, it is recommended that all casks follow the same route between the decontamination area and the loading area in the spent fuel pool. Strict supervisory con-trols will be used to restrict the casks to the path marked in Fig. 9-2. The only exception would be the fuel transfer cask, which m'ust occasionally be moved over the reactor. Its path is shown as the broken line in Fig. 9-2. / The casks being moved at elevation 633'-0" should be moved I at approximately 6 inches above the floor and would only be kg,,nearedgedropsattwolocations. The first location is near the decontamination area where a drop or tumble would not en-danger any safety-related equipment. The other location is at _, ) the southwest cornerfof the spent fuel pool which,has been modified to withstand the consequences of a cack drop. The safety cables will always be attached to the fuel transfer cask, even when not moved over the reactor. /' Figure 9-2 also indicates the area to be used for decontami-nating the cobalt cask and TREAT II cask. This is illustrated on the 632 '-6" elevation and to the west of the spent fuel pool. This area as well as the path taken between the spent fuel pool and the decontamination area above room 44. will be painted to guide the cask movements en* route through the containment building. 9.4 Procedural Modification Tes ting, inspection, maintenance, and handling programs are already in effect. These programs shall be, at least, in ac-cordance with the following procedures. 9.4.1 Testing I i A test program shall be established to demonstrate that the equipment will perform satisfactorily in 9-3

service. The testing shall be performed in accordance with written test procedures which incorporate the re- ) quirements and acceotance criteria contained in appli-cable design documents. Test results shall be documented, and evaluated by a responsible authority to verify that test requirements have been satisfied. Testing as de-tailed herein involves operational tests. These opera-tional type tests cover checks of control functions and capabilities. 9.4.2 Inspection Handling equipment subject to inspection prior to handling shall be listed in an inspection procedure. Inspections as detailed herein include three types: frequent, periodic, and major, and apply to equipment in use. i g-Evidence of inspections, and the results of periodic and 6 major inspections and follow-through shall be made avail-able. 9.4.2.1 Frequent inspections are those performed on a day-to-day or similarly frequent basis. The inspection shall con-form to consensus standards and federal, state, and local health and safety regulations. The inspection coverage shall include parts essential to safe opera-tion and those parts recommended by the manufacture-A checklist shall be used to perform the inspections. These inspections shall be performed by the individual responsible for the operation of the particular equip-ment or by another competent individual. J 9-4

9.4.2.2 )' Periodic inspections are those performed on a preset in-terval. The inspection shall conform to consensus standards and federal, state, and local health and safety regulations. The inspection coverage shall include parts essential to safe operation and those parts recommended by the manu-facturer'. Periodic inspections shall be made by an employee or agent responsible for the operations of the system or component who has been qualified by experience or special training to perform such inspections. Results of periodic inspections shall be documented. 9.4.2.3 Major inspections are those performed on an annual basis and shall conform to a prepared procedure. Inspection T~ s coverage shall include recommendations of the manufacturer '~# or designer and shall conform to consensus standards and fedetal, state, and local health and safety regulations. Visual examinations or nondestructive examinations shall ~ .be used for these inspections. Results of major inspections shall be documented. 9.4.3 Maintenance A maintenance program shall be established to ensure that the handling equipment is maintained in good operating condition. 9.4.3.1 Equipment shall be serviced at specified intervals in () / accordance with the manufacturer's recommendations, severity of service, and environment. Items damaged or 9-5 i

worn sufficiently to affect operation of the equipment shall be repaired or replaced before continuing opera-tions. Replacement parts shall meet or exceed the specifications of the part being replaced. 9.4.3.2 Maintena'nce shall be documented and the records kept current. These records shall show lubrication, ser-vicing, adjustments, repairs and replacement of the equipment. 9.4.4 Handling This section contains requirements to be fulfilled by the individuals who will have operational control of the handling equipment in use at a nuclear power plant. 9.4.4.1

T Handling and moving clearances shall be investigated.

Evidence of maintenance in accordance with Subsection 9.4.3 shall be verified. 9.4.4.2 The handling of casks shall be in accordance with detailed, written, approved procedures, instructions, or drawings. The procedures shall include the following as a minimum: a. Responsibilities shall be defined for key individuals. b. Equipment to be used shall be identified and its se-lection shall be on the basis of its capability to handle the load. Loads handled shall not exceed the loads used in the design of the equipment. c. Manufacturer's operating instructions and conditions h of operation shall be followed. I 9-6 y -w-r

d. Work instructions for single tasks that, because of their relationship to each other, must be acccmplished k'" in a certain sequence, shall be issued. e. Where applicable, acceptance criteria shall be speci-fled to be used as a basis for determining when a task has been satisfactorily completed. f. Inspections or check points which require documentation by specific individuals as proof of satisfactory com-pletion shall be included. g. Procedures shall identify maximum safe loads which are permissible, and shall describe specific methods of ensuring that these safe loads are never exceeded. Load indicating devices, properly calibrated, shall be used in systems where the primary source of power has the capability of imposing excessive loads to the equip-ment, component or item being handled. s 4 e 1'

l 9-7

L u u u u c u __. - - y' i l'Ca t;~c"EC 2 C -- %N Shipping Cask Area - I CCCC 'Ecc c5cS )m C CC CQ!-C CC C C C QDD n n t, C OC C C C,CC, C rim rn e r /h5N055 Spent Fuel Storage L; _ _ _, u ; Racks 7 "~r'" nE D No::le Rack p hnuawc g -nr .- 7 L-h C L % rmaam- ,- O a c7: ";II-1tI SupportTubesand,wp;hc;t Rack for Guide 9 gti Channels ~ - :.y P / Plan Spent Fuel Storage Pool (Present) No::le Rack ):,, \\_ = Cask Area g_ m I pact Pad 3 i y._. ..JJULLuuL.u u' J 2] CCE'".CO C C C: 23pGth2C CCO C-O C CL C C'O D li I JOCCOOCCGCl 'c ]0 E0 O n E 0 l EUJWLlLi D"EC C Ci t GJuJ'a GG.Tbb, .- x DCEfOU C! DOCO2C n m n n '.- Plan Spent Fuel Storage Pool (Proposed) )(./ FIGURE 9-1 Plan Of Spent Fuel Storage Pool And Proposed !!cdification j

[ tia,CTC4 Cut Q4N3 )sta;ct,vtSSEL /.= \\ ~ .. -.-N.

%,,,, N'N..

QEA.CTCQ But3NG r Ecurca l I i t.- y ./,. ,I - m 3. - utt,. a (=m s. gs l O, K i i. / f t,,.. //

e. m.i r ;,

.,. g. y' ,,,L 's ~, 4f 54,(r ii, j'. surmont / l. 4 / ' 91 11 a T2 tut Postric e ce coa +it i, !l \\ a l

e. 6n.;-' u ll /

) i !f, ',, lik---[ ,,.{g,,,_) 1 \\ um -; / [idA ^k /fr-- \\' ' ^*' " ' d / = 4 = ~.. .+ w ,3*M -mm .f. 23 .\\ 1-3 \\ , " w a> ,cs emm ..).... e i L. l " a::.* l .j .f.. w;.'. li /. 4 I' .s'i l i e tt. . \\, l l

l i

'II. v e m l p v. un.s I Vw i ll r-. -..4- \\ *' J i

j.---

!!ii N N j I tut ct ev e; . l8 f.l g,~ ca.:ew p g k I l / /l ] G-. - r-!,I, I / 4-+ e jl / \\gI j 3 .t e te ii a r - c4- .i at - i \\ l l [.,- t. 'e N..

) v%

i y" ""j jj ,' ~ ' 5~ j h y}'n %e g en g Decon w, - Zone g '. ' ,1 g u t=% i \\ \\. d_s i.-'.. g irr :: l l.M,,.:in. s \\\\ \\ p -j -~c ./ ,/ .1 ~

i

_._.{ -: ;p. !.. f,,e u

4.....

g ./ = .\\ '+ a,i#ib, s % N,:!i l- .S / ? s. .h.tt43:.cj; I .'g \\ ~ lI - /\\l.:t h I .\\ 3 i V :.A *-'; " d-2 7t;", g l [ ~- 8 P

spac: Pad 4

ws< / u.tsi 57 / \\ %, Nik itaanste. cats... g

a

,s niwg'.( ) / 4,e j s .\\ i wI 4G.N ~ s em / . uf s, it, 612'. 4* N. N U-L3 t ,a t I l l FIGURE 9-2 Proposed Cask Routing i

Y% %s 8 ,4 + i e { LIFTING LUGS 3'2 CARB06: STEEL DISTRIBUTION PL. '4" STAINLESS STEEL BOX

{

i ) t}r'l

1 !f I

a - "M .I }

_,lg ki y!b l

l lil 0 [ ? r i i r -e (IIEXCELL I. ~' 4 -5052.003 S ECTION A ALUMINUM llONEYCOMD ..I\\ , l'-3 }" l'-33 " 4' ou = ; (Typ) TYR c H U R R H H h 2" STAINLESS I STEEL ROD e d e ,a _7 A _F. I. 'r u u a a u u __ e _. O, LIFTING LUG 6'- 7'2" (TYP) (Typ) 19'-10 2 = PLAN FIGURE 9-3 Impact Pads W h b

l l

10.0 REFERENCES

h 1. Bechtel Power Corporation, " Design of Structures for Missile Impact", Topical Report BC-TOP-9, Revision 1, July, 1973. 2. Johansen, K. W., " Yield Line Theory", London, 1972. 3. Timoshenko, S. and Woinowsky-Krieger, S., " Theory of Plates and Shells", McGraw Hill, 1959. 4. Amendment No. 10 to Final Hazards Summary Report for Big Rock Point Plant, Vol. 1, November, 1961. 4 O e f

t...

10-1 e m

g 8 APPENDL. A SAMPLE CALCULATIONS 4 e s M

CASK DROP ACCIDENT SAMPLE CALCULATIONS _ 'f The calculations presented in the following are based on the procedures outlined in Ref. (1). Detailed Calculations: See paragraph 5.2.1, Cask Drop Accident Case (1). Cask: Fuel t'ransfer cask Drop Location: At or near center of spent fuel pool, from eleva-tion 633'-6", one foot above the top of the refueling floor. Near the center of the spent fuel pool slab at elsvation Target: 602'-1", 7 ft. from the pool south wall. Drop height in air (S)......... 3'-0" Drop height in water (H)....... 28'-5' Total drop height (S) + (H)...... 31'-5" A._. Weight of loaded cask (W,) ..... 48.0 kips Total length of cask (L)....... 14'-8" Base diameter of cask (D)...... 3'-0" 2 Maximum cross sectional ' area (A).7.07 in 3 Weight density of cask (y,).... 605 lbs/ft 3 Weight density of water (y).... 62.4 lbs/ft 2 Gravitational acceleration (g). 32.2 ft/sec Drag coefficient (Ref. 1) (C ). 0.85 D Velocity of cask when reaching water: v, = / 2 (g) (S) ..................................(Eq. A-1) V, = /2 (32. 2) (3. 0),........ 1).9 ft/sec s _. A-1 l # l l

6 Calculate velocity of cask at impact - use formula derived ) from equation D-7, Ref. l.: V2 _ v22 -2aB ................................(Eq. A-2) =e V2 _ v2 o 3 where: (C )(y)(A) d ................................(Eq. A-3) a= 2 04o) (0. 8 5) (62. 4) (7. 07) 0.00391 ft1 = a= 2(48,000) 2 2 193.2 ft /sec V2 2 (g) (S) = = o Vj=1(1-y/y,] .......................(Eq. A-4) Vj = gy_62 4]= ... 7387 ft /sec .2 2 2 g m(_. Vj = E [1-y/y,(1+aL) ] ..........................(Eq. A-5) Vj=0bO391 [1-j (1+. 00391 (14. 8 ) ) ] ; 2 2 Vj =....................... 7,337 ft /sec -2aB -2 (. 00 3 91) ( 2 8. 413 ) =.. 0.801 i g =e v2. y22+e-2aH(y2_y2); o 3 V2 = 7387 + 0.801[193.2-7337]; 2 2 V2 = 7387 - 5722 =.......... 1665 ft /sec V = velocity at impact =.... 40.8 ft/sec nf g-A-2

4 KINETIC ENERGY OF CASK AT IMPACT: W (V) 2 h ...............'......................(Eq. A-6) K.E., = (g) .0(1665) =.... 1241 k-ft K.E. = o 2(32.2) FLOOR SLAB, DIMENSION,' MASS, ENERGY: Floor slab total thickness........... 6'-7" Structural slab thickness (t)........ 6'-0" Long span............................ 26'-0" Short span........................... 20'-0" Target center distance frca south wall 7 '-0" Effective mass of floor: W*

where

M =- e 9 sf ' ) [ (6. 0) (10.17 ) 2+ (0. 5 8 4 ) (3. 5 84 ) 2 ] [0.150]( Ns W = e ............ 74 h W = e ~ PLASTIC 3MPACT ENERGY REDUCTION FACTOR: M W 48.0 = 0*393 A = M +M W +W 4 8. 0 + 7 4. 0 = = o e o e where M, is mass of cask, M, effective mass of structure. Energy transmitted to floor: K.E. = A(KE ) ..................................(Eq. A-7) g K.E. = 0.393 (1241) = 487.7 k-ft FLOOR RESISTANCE AT YIELD (REF. 1,2) ,n - 2r(1+1)M 34.., R ................................(Eq. A-8) = 2 1 y(_r ) o 1 l l A-3

. whors (i) the degree of rostraint ist M' A' i=N-"% ( 8); (see Ref. 2) "s u / r = cask base radius o r = least distance to support y M = slab moment capacity at yie3.d u s1(0.9) (f ) (DIF) (d "2I) +A' (0. 9) (f ) (DIF) (d 'd)... (Eq. A-9) M =A s y y u A (0.9) (f ) (DIF) ...........................(Eq. A-10) where a' = 0 85 (f') (b) C For this example: Thickness of the structural slab is: 72 in t =.............................. 3 3 d =72 7 - lg =................. 69.9 in 1 12 in b =.............................. 2 1.56 in fge A, = (#11@l2" E.W.) 2 0.60 in /ft (#7012" E.W.) A' = (- sf' =.............................. 3.0 k/in2 2 (40.0)(0.9)................... 36.0 k/in f' = y 1.20 DIF =............................. l 1 I (0. 9 6 ) (3 6. 0 ) (1. 2) 1.36 in a, = = (0. 8 5 ) (3. 0 ) (12) 2 0.96 in /ft; i A = A - A' = i si s s l l Then: (0.96)(36.0)(1.2)(69.9 1*36)/12 M = 2 u .+ (0. 6 0 ) (3 6. 0) (1. 2) ( 6 9. 9-1. 0 ) /12 3 386.5 k l O*60 i = 1.56 = 0.384 l bE. (3.0) = 0.214 r 2(7.0) 1 [2n] (1+0. 38 4) (3 86.5) }l' R= = 3,922 k l l 2 1-3 (0.214) A-4 .. ~.

.. SLAB DISPLACEMENT AT YIELD (REF. 1; TABLE 4-3) : (a) (R) (a) 2 (1_y2 ) .......................... ( E q. A-l l ) X = 7 12 (E) (I ) a where: 1 .................................(Eq. A-12, z, = 1(1,.1,,) z,=iz(e)3(b> .................................(Ee. A-13) I = F(d) 3 (b) ..................................(Eq. A-14) er For this example: = b- (72) 3 (12) = 373,250 in"/ft; I g 12 (0.012)(69.9)3(12) = 49,180 in"/ft; I = I = b (323,250+49,180) = 236,210 in /ft; 4 a 2 2 Concrete modulus of elasticity (E)..... 3,000 k/in )'( Concrete Poissons ratio (V)............ 0.17 Displacement coefficient (c)........... 0.0803 See Ref. 1; Table 4-3 Displacement at yield (Eq. A-11) : 2 (0.0803)(3922)(240)2(1-0.17 ) 0.025 in = X = y (3,000)(236,210) EFFECT OF DEAD AND LIVE LOAD: q = 0.15 (6. 58) +0.0 6 24 (28. 41) = 2.77 k/ft2; D BENDING MOMENT AT CENTER OF SLAB (Ref. 3; Table 30) : b 26 M = 0.0327q " ' II # - " Yg = 1.3) D D M = 0.0327 (2.77) (20.0) 2 = 36.0 k D A-5

I Displacement at center of slab: }~ 0.0209qD ^I I = XD E(t)3 0.0209(2.77)(20.0)"(12)2 0.0012 lb/in = X = D (3000)(6.0)3(12)3 Resistance of floor slab (after subtracting for dead and live load): M -M R' = R( u D) 3 u 3922( 386.5-36.0); R' = 386.5 R' = 3556 k l i ENERGY-BALANCE: X'l KE = R'(X' y-) ~. J' '

(4 87. 7) (12) = 3556(X' 0.025.0012);

12 N ' b. L, I487*7II12I = 1.657 in; 0.012 + X' = 6 TOTAL DISPLACEMENT: X = X'+X = 1.657 +.0012 % 1.66 in: D MAXIMUM ALLOWAELE DUCTILITY RATIO (REF. 1: TABLE 4-4): 10 , 10(12)(69.9) = 87 > 30' p = 30 y p-p' (1.56-0.60)(100) ALLOWA9LE DISPLACEMENT (REF. 1): 30 (0.025) = 0.75 in X = p(X ) = y d' The total displacement X = 1.66 in exceeds the allowable dis-A# placement criteria. A-6

y . o _.. _ 1 NOTATIONS I-The notations are defined in Ref. (1) or explained in the sample calculations. Terms not defined or explained in the calculations are listed below: A ' Area of cask cross section. A ...A' Area of rebar. s s a' Depth of stress block. b Width of stress block. d Depth of slab. DIF Dynamic Increase Factor. E Elastic modulus. f' Concrete strength. c f Rebar yield stress. y 1 Degree of restraint. M Dead load moment D ) M Ultimate Moment g p...p' Percentage rebar. q Dead load. p Impacted area radii. r,...r7 S Drop height in air, t Thickness of slab. A Plastic impact energy reduction factor. l I i j e e f A-7 l

Po l 1 2 1 APPENDIX B ~ TESTS OF EQUIPMCIT AND CCMPONENTS USED FOR HANDLING CASKS 9 O l' I t i I i

7 o-r. -) I. ACCEPTANCE TESTS The following is an outline description of the tests that have been performed in accordance with accepted code pro-cedures. A. Report On Acceptance Test No. 7 - Fuel Handling System The acceptance tests for the cranes and the trolley hoist for the fuel handling system were essentially completed on September 19, 1962. The following observations and data were recorded: (1) Reactor crane speeds were measured with stop watch and tape measure, and the speeds were compared with those listed in the crane instruction book. The latter are indicated in parenthesis. f Bridge Speed: 53.5 & 1.18 fpm (50 & 1.2.fpm). Trolley Speed: 55.5 & l.18 fpm (50 & l.2 fpm). 75 ton Hook Speed: 8 fpm, both up and down (6. 5 fpm). (2) The fuel transfer cask was rigged, using the slings provided for the vessel head. A third sling is re-quired to balance the weight of the hoist mounted on the cask. Speeds were again checked while carrying the cask and were found to be essentially the same as those in paragraph (1). (3) The fuel transfer cask was lowered into the reactor vessel to check the lower limit. It stopped when the bottom of the cask was approximately 103 inches below the top of the shield tank. B-1

. ~ sl '1 r (4) The safety sling was checked by suspending the coffin from the 75 ton hook, tripping the wedge ) assemblies, and then lowering the hook until all the weight was supported by the safety slings. The load lowered 2 inches before the wedges engaged. (5) The~ crane interlocks were tested, both with the reactor mode switch in the REFUEL position and with links open to drop out relay 4k30B. Power to the crane was interrupted when the mode switch was placed in the REFUEL position, and it was re-stored when the key interlock switch was used. The links were opened to check the dead band area over the reactor. The dead band was found to be as described in the test procedure. II. OPERATIONAL TESTS AND INSPECTIONS D A. The Whiting Corporation conducted inspections of the two cranes (S/N8654 and S/N8577) during the 1960's and on October 1, 1971. B. The cables for the cranes are tested and inspected in accordance with Maintenance Procedure MFHS-2. C. The fuel handling equipment is tested and inspected in accordance with Operating Procedures for NFS-100, Shipping Cask - Fuel Handling Equipment Safety Check - Test No. TR-07 and Test No. TR-02. D. The hoist and cranes are inspected and tested in accordance with Procedure No MCLP-1, Hoist / Crane Inspection. This procedure defines steps to be taken to comply with the regulations of the Michigan i Department of Labor, occupational Safety Standards g B-2}}

ML19305B749

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ML19305B749

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