ML20079C402

Analysis & Safety Evaluation of Spent Fuel Shipping Cask Handling at Monticello Nuclear Generating Plant

ML20079C402
Person / Time
Site:	Monticello
Issue date:	01/13/1976
From:	Salewske M, Schwartz K, Silvester A NUCLEAR SERVICES CORP.
To:
Shared Package
ML20079A881	List: ML20079A880 ML20079C402
References
LS&R-NOR-0150-0, LS&R-NOR-0150-06, LS&R-NOR-150, LS&R-NOR-150-6, NUDOCS 9106190400
Download: ML20079C402 (61)

Text

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Licensing Report LS&R-NOR-0150-06 t

January 13, 1976 Jtegulatory Docket Blt.

MMM,MfLQ AN ANALYSIS AND SAFETY EVALVATION OF SPENT FUEL SHIPPING CASK HANDLING AT THE MONTICELLO NUCLEAR GENERATING PLANT e

Prepared for NORTHERN STATES POWER COMPANY by NUCLEAR SERVICES CORPORATION Campbell, California 95008 Prepared by: .g Reviewed by:

M. Salewske A. Silvester i

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Approved by: (24 K. Schwartz uber L. Stanley Project Engineer ManagerofPr[ojects i

i 9106190400 760122 PDR ADDCK 05000263 P PDR

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t T ABLE OF CONTE'<TS PAGE.

1.0 INT RODUCT10ti 1-1 2.0

SUMMARY

AND CONCLUSIONS 2-1 3.0 STRLCTURAL ANALYSIS Of SPENT FUEL CASK IMPACT 3-1 3.1 Criteria and Bases for Structural Analysis 3-2 3.< Description of Structures 33 3.3 Loading Conditions 3-5 3.4 Analytical Procedure 35 3.E Results and Conclusions 3-11 4.0 EV?tUATION OF CASK HANDLING SYSTEMS 4-1 4.1 Crane System Safety Margins 4 -1 4.2 Crane System failure Analysis 4-1 4,3 Crane System Maintenance 4-3 5.0 EVALJATION OF PLAfti S AFETY FE ATURES 51 5,1 Integrity of Spent fuel Pnol 5-1 5.2 Protection of Spent fuel 5-2 5.3 Protection of Vital Equipment 5-5 C." ADvl'.1STRATIVE CONTROL OF CASV HANDLING OPERATIONS f. - l f.1 Restricted Cask t'overient rath 6-1 t.2 Crane Systom Pre-Lift Checkout 0 -2 5.2 ;'rocedures 'or Pet eiving leiding and Shipping C.3

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, :iF:H 4CES /-l dPPE'D:/^ Lettription of N!S-4 Cad scr ,s;m / r 3 31,wr W de and Fffect- Tnelysis i

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kI S.T_OF, FIGURE S 3-1 Structural Framing Plan for the Operating Floor at Elevation 1027'-8" J-2 Structural framing Plan for the Spent fuel Storage Pool Slab at Elevation 988'-11" 3-3 Structural framing Plan for the Cask Receiving Area at Elevation 935'-0" 34 Structural failure Mechanisms for Beams 3-5 Energy Absorption Capability for Steel Beams 3-6 Mechanisn for Small Height of Drop at Edge of Slab 4-1 Monticello Reactor Building Crane Operating Envelope 5-1 Protective Arrangement for Spent Fuel Pool 5-2 Reactor Building Seismic Analysis Acceleration Diagram (N-S Direction per the Monticello FSAR) l S-3 Reactor Building Seismic Analysis Aereleration Diagram (E-W Direction per the Monticello FSAR) l 6' Sacrt Fuel Cask Movement Path l

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A1 !bF ; " cent Fuel Shipping Cask i

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l.0 INTRODUCTION Northern States Power Company has contracted Nuclear Services Corporation to provide an engineering safety and licensing evaluation in response to Nuclear Regulatory Comission (NRC) questions en the cask drop accident for the '

Monticello Nuclear Generating Plant (Reference 1). Northern States Power Company has previously responsed to these questions in Octobe.',1975 (Refer-ence 2). The emphasis during this previous evaluation was placed upon the General Electric Company IF-300 spent fuel shipping cask design (Reference 3) ahich was te be used to ship spent fuel from Monticello in 1976. Since pre-liminary results of this evaluation indicated that the reactor building struc-tures cannot, for all cases considered, withstand the impact of a dropped IF-300 cask, Northern States Power Company has advised the NRC of both interim and long rance urograms for renoval of spent fuel.

Long range planning has been initiated as described in a status report to NRC in May, 1975 (Reference 4). However, due to the length of time estimated to imple-ment proposed plant modifications to provide redundant cask handling equipment at Monticello, an interim program for spent fuel shipping, using a lighter two-  !

element shioping cask, has been established by Northern States Power Company to acconmodate anticipated spent fuel shipments in the near future. In accordance with the Reference 4 response to NRC, a safety evaluation for a lighter two-element shirping cask has bee's completed using the Nuclear fuel Services, Model NFS-4. SHrcino cask (Reference 5). Safety aspects evaluated include:

1 Reacter Building Structural Capability

?. Cask Handling Systems Design Integrity Drotection of Spent fuel and Vital Systems

. Adninistrative Control of Cask Handling Operations

! Safety evaluation results are documented in Sections 2,0 through 6.0 of this re-l port. Thcse results provide the necessary engineering, operating and licensing l information for a favorable resolution of the cask drop question for the 25 ton NFS-4 shioping cask at Monticello. Further evaluation is being considered to i

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extend this program to fuel shipnino with larger casks such as th,e General Electric if-300 cask, faveioble results for this evaluation would provide I a more tinely solution to spent fuel storage space problems at Monticello.

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l 0. SUWARY AND CONCl.USIONS i

An evaluation of plant safety during the handling of spent fuel with an irradiated fuel shipping cask has been completed for the Monticello Nuclear Generating ilant. The evaluation results are detailed in the following sec.

tions of the report.

Safety evaludtion results are based upen engineering analyses of the Monticello reacter building structures 'or the ef fects of a NFS-4 two-element ( Appendix A),

spent fuel shipping cask drop accident. The safe lif ting height for the cask abcVe the operating floor was deterrined and the effects of dropping the cask over the spent fuel storage pool and equipment hatch were investigated.

Analysis results indicate thit a six inch safe lif t height is pennissible for the operning floor and that no losa of structural integrity would be experi-c enced by the spent fuel pool A safety analysis, using the single failure criterion, has been conducted on the react :r building 85 ton bridae crane, The crane system is not designed to be single f ailure proof and may be subjected to either single random mechanical or conteol system failures. There are no single random electrical failures that cat. ret be c moensated for by operator action to prevent an uncontrolled load irocring. Tor those mechanical or s'.ructural failures which could result in a loo- Jrrt , very large (greater than 10) ' actors of safety exist in the crane

.oC anen ti fer the handling of a 25 tc, shipping cask. Furthermore, preventa-t',' ": 'n'erince. inspection and opera'.ional control requirenents will be l

ir~ w , .ee

• c ver'*v the ir*.egri*y of the hoist sys tem, control systems and i

cuer n;1i v v t5e cr ane.

Int 4 ' * - / i.' DI W safety featdres he been assured by structural analyscs and in ir%e entation of administrative controls for cask handling. The spent fuel f

pool will ntt be seriously damaged by cask impact and, therefore, only a minimal l

loss of aner from the fuel pool must be considered. Spent fuel stored in the fuel pool will be protected from damage by separation of fuel and cossible cask drop locations during cask handling operations over the pool. Thus, spent fuel will not be vulnerable to a cask tipover or vertical drop condition. Vital plant equipment contained below the cask receiving area and equipment hatch will be 1 -

k protected from darra.,1 N 'rt'rictinq the cask hoist position to a prescribed locatior.

Administrative control of cask and fuel handling operations will be developed in accordince with the Monticello Technical Specifications and the Northern States Power Company Quality Assurance Program. A strictly enforced cask travel path will be employed as well as detailed procedures for reactor building crane pre-lif t checkout, cask receiving, cask loading and cask shipping.

Based upnr the foregoing safety evaluation results, it is concluded that satis-factory . ant and operational safety, during spent fuel and cask hanaling at Yon *iccil' ,,ill be maintained. The results also provide the neces w y infor-

at'r fo' regulatory acceptarce of the NFS-4 shipping cask at Monticello, 22-l

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3.0 S'i.ivCTUDAL ANALYSIS OF SPENT FUEL CASK IMPACT Engineering analyses have been conducted to predict the effects of a cask drop upon plant structures during handling over the spent fuel pool, while being moved across operating floor areas, and while being hoisted in building access and hatch openings. The results obtained from the structural analyses give the maximun, urop neight capability for a 50,000 pound fuel cask (Reference 5) at four locations on the operating floor, two locations on the spent fuel pool floor, and two positions down the reactor building equipment hatch.

For each drop location, the principles of conservation of energy and conserva-tion of linear momentum have been used to evaluate the effects of impact on slabs and beams for flexure, direct shear or punching shear, in evaluating the cask impact on the spent fuel pool floor, effects of bouyant and drag forces on the velocity of the cask at impact were included. The analysis employs the following techniques which are in conformance with regulatory criteria estab-lished by References 6 and 7; this includes:

(1) the yield line theory for the collapse mechanism of concrete slabs; (2) plastic analysis for beams and the utilization of the inherent ductility present in slabs and beams beyond their yield loads to absorb the kinetic energy of impact; (3) energy absorption by the defonnation of the cask impact devices; (4) the increase in allowable stresses in concrete and steel to account for the dynamic nature of loads; and j (5) the resistance to direct shear in slabs due to available shear t

friction reinforcement.

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3.1 CRITERIA . .) ..sr.! FOR STRUCTUF1.L ANALYSIS The following areas rf "r " actor ouilding were investigated:

e The operating floor at elevation 1027'-8" e The spent fuel storage poci at elevation 988'-11" e The cask receiving area at elevation 935'-0" (below equipnent hatch).

The principles of conservation of energy and conservation of linear momentum nava been employed to evaluate the energy transferred to slabs and beams during the cat.k irsac t. The internal energy, or the energy that can be absorbed by slabs are beams, is determined by the plastic or limit analysis theory which takes into account the inherent ductility present in the structural members.

Th? force acting on the target structures during the cask impact is determined by considering deformation of fixed impact limiters attached to the cask, penetration of the cask into the target structures and deflection of the target structures.

r ostu'ated drops of the cask over operating floor and equipment hatch areas were e

2sune i to result in a free vertical fall with the cask bottom end impacting the crgat ;;ructure. For a drco into the spent fuel storage pool, it was assumed that eithe a vertical drop occurs into the water or that the cask hits the

-40 M t9? ico', tips-over and drops horizontally into the fuel pool water and 71_ ; . tu .301 slab. This assumption is highly censervative since cask tip-ove' is nc.t e.eccted to occur as discussed in Section 5.2. However, both 90eizcit3- end vertical drope of the rask were considered for the evaluation of t ^ r. , f. - - ':.el r uol s tr ucture .

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i The following material properties and structural assumptions were used in the analysis:

(1) The compressive strength of concrete, f' = 4000 psi.

c (2) The tensile strength of steel reinforcement, f y= 60,000 psi (3) The yield strength of structural steel, f = 36,000 psi (4) Weight of concrete is 150 lbs/ft 3 (5) Nominal live load for the operating floor is 100 lbs/ft (6) Impact load is distributed uniformly over the area of cask impact.

(7) Metal decking for concrete slabs is neglected.

(8) The spent fuel pool steel liner plate and one inch setdown plate is neglected.

(9) The mitigating effect of empty control rod and spent fuel racks (installed on the pool bottom) and their crushing strength is ignored.

(10) Allowable stress of the materials is increased by using a dynamic increase f actor as recommended in Reference 8.

3.2 DESCRIPTION

OF STRUCTURES The cask will be lifted through the equipment hatch located between column rows R, S, and 6.9 and 7.9. The cask will be hoisted to the operating floor level at elevation 1027'-8", and then moved to the cask laydown area before lowering it into the spent fuel pool. The movenent path of the cask will be controlled based upon the results for the impact analysis locations indicated in Figure 3-1 (see Section 6.0). Based upon the proposed path and the structure design, a

'otal of eight critical cask impact locations were selected for detailed analysis to determine the effects of a postulated spent fuel cask drop upon the existing plant structures. Locations selected were (Figures 3-1, 3-2, and 3-3):

e Four (No. 1 to No. 4) on the operating floor at elevation 1027'-8".

e Two (No. 5 and No. 6) on the spent fuel storage pool slab at elevation 988'-11".

e Two (No. 7 and No. 8) at the base of the equipment hatch at elevation 935'-0".

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e locations No. I and No. i' cie at the mid-span and edge of the operating floor slab, respectively. The slab has a nominal thickness of 13.5 inches of concrete 4

over 18 oage metal decking which spans between steel beams, and has steel reinforcement at top and bottom in both directions. To be conservative, the slab with the least reinforcement (No. 6 reinforcing bars spaced at 12 inches on center, top and bottom running in both directions) is considered and analyzed as a one way slab in the analysis. Location No. 3 is at or near the center of steel beam W 21 x 55, which is the lightest beam in the area of cask movement.

Since W 21 x 55 is connected to the floor slab with 3/4 inch diameter shear connectors, it is treater as a corposite beam in the analysis. Location No. 4 is (elected at or near the center of steel beam W 36 x 160 adjacent to the east eage cf the opening for the equipment hatch (Figure 3-1), The beam is considered as an ordinary Steel beam in the analysis.

The spent fuel pool floor slab, where locations No. 5 and No. 6 are chosen (Figure 3-M, is reinforced concrete accroximately 6'-b" thick. The slab is more hcavi'y reinforced in the east-west direction than in the north-south direction. Steel beams in the east-west direction are fully encased in the corr ete, and form an integral part o' the fuel pool slab. The slab is s accrted ne fcur sides by deep concrete girders. The water above the pool

('at is acma!'y cortrolled at approxirately 37'-9" in depth, in the event that tir % 3ter 'evel rea:hes the it.inimum allowable Technical Specification depth of 33'-0", t'e additional free fall of the cask (57 inches) would slightly

'r rem " . velocity at impcct with the slab. This increase in the energy

'o w a* scried wou'd reduce the factor of safety (h /h,) computed by' approxi-cataly M Hee Section 3.4).

Tt: fire S A et the .use of the equirent hatch at elevation 935'-0" (rig:;re 3-3) 'l cornowd o' two dif ferent types of construction and is separ-ti t ed by three 'o0+ ' hick concrete wall running diagonally underneath it.

Location 'io. 715 chosen near the east edge of the equipment hatch, where the floor slab is reinforced conc. rete approximately 24 inches thick with encased steel belms. Location No. 8 is at the intersection of the three foot concrete 4 -

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wall and a 68 inch tn'rt, '.oncrete lab which is composed of pour-in-place

.oncrete over precast rein 0 rced tuncrete inverted T-beams. The three foot l thick diagonal concrete wall, which sits directly on the base slab of the i reactor building, is very stiff and can safely withstand the cask drop from the ,

maximum li f t height. Therefore, a detailed analysis need not be conducted for the concrete wall.

3.3 LOADING CONDITIONS In addition to the cask impact load, the dead weight of the structure, live load on the tlcor and weight of water above the fuel pool floor were included in the analysis. Since the impact load depends upor the weight and drop height'of the cask, the effective mass and stiffness of the target structure and the eigidity of the cask itself, it was determined individually for each impact location.

The analytical procedures described in Section 3.4 were used to detemile the impact load. Based or the relative position of eich cask impact location with respect to the tatget structure, the following two different cases were considered:

o Central Impact - cask impact locatit is at or near the center of slabs and beams.

e Edge impact - cask impact location is at the edge of slabs, tne distance between the edge of slab st;oport and the impact location is less than the effective depth of the slab.

Foi postulated drops into the spent fuel storage pool, the effects of buoyancy and drao forces on the velocity of the cask at impact were also included in the melys" as developed in Reference 8.

3.4 ANALYTICAL PROCEDURE The impact between the cask and structure was considered plastic due to the local perrenent deformation either to the cask or the structure, or both.

Therefore, a considerable amount of energy is absorbed or dissipated during the impact. The remaining energy is transferred to the structure (Reference 8 and 9). The structure will survive the impact only if its energy absorbing capacity l ,

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i is equal to, or larger than the r qy transferred to it during the impact.

Besed on this principle, the energy balance rethod was used to detertnine effects of cask drop on slabs and beams. The analytical procedules are described below 'or each energy 'bing mechanism.

3.4.1 Ene_rgy_ Transferred to _the___ Target Str_ucture The kinetic energy of the dropped tvsk, just before the impact, is given by:

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Ea fM)V)2 whe re :

M)

• Mass of the cask V) = Velocity of the cask at impact, which is equal to hh for a free drop (g is acceleration of gravity, h is height of free drop).

Equating lineat momentums before and after the impact gives:

M)V) = (M) + M2)V2 (2) whe re :

92 = Effective mass of the target structure, depending upon the type of impact considered.

V2

• Velocity of the cask and af fected structure af ter the impac>,.

The Lin* tic erergy of the cask and the af fected structure af ter the plastic imoset is giver by:

E e " f- (M; + M 2)Y2 (3)

Substituting for V 2f rom equation (2) gives:

M a" IF~1 T T )M 2

Y11 (4) 6 -

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. l E, is the energy transferred to the target structure, or equivalently, the energy required to be absorbed by the target structure. Reference 8 presents a formulation to obtain the effective mass (M 2) of the target slabs and beems.

In the case of cask impact at the edge of a slab, the effective mass of the l target structure is very small. In this case. the inass ratio M)/(M) + Mg ) is  !

conserystively assumed to be 1.0 (Reference 9),

i 3.4.2 Energ,y Absorbed by the Targe,t_$_t_ructure The amount of energy that can be absorbed by the target structure depends not only on the properties of the structural material, but also on the boundary and loading conditions of the structure. To cover the eight different locations selected for analysis, two different cases must be considered for slabs and ,

heams: central impact for slabs and beams, and edge impact for slabs. t 0

3.4.2.1 Central _ Impact for Slabs and Beams One Way Concreto Slabs and Composite Beams A one way slab can be treated as a beam by assuming that the effective width of the slab during impact equals two times the slab thickness plus the diameter or length of the cask impact area. The energy that can be absorbed by the beams can be found based on the boundary conditions and failure mechanism. The boundary conditions are assumed to be fixed if the slab is continuous or i rigidly connected at its support, otherwise a simply supported beam is assumed.

The three hinged mechanism and the one hinged mechanism shown in figure 3-4 were considered, respectively, for a fix-ended beam and a simple ocam to obtain its energy absorbing capacity. I The rotation. Otu, due to the ultimate moment uM acting at a hinge location is calculated using the method given by Mattock (Reference 10) and Corley

, (Reference 11). These parameters are used to obtain the energy the beam can absorb (E4 ) in rotation.

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Therefore, Ii U I6) ui tui i

Eque. ting jE and E, from equations (4) and (5), plus tne work done by the existing loads gives the maximum allowabls drop height (h ) that the slab or beam will m

sustain if a drop is assumed. A factor of safety is obtained by dividing h by the contrulled lif t height (h )s above the target structure.

Total deflection (6 t) of the beam equals the sumation of the elastic deflection and plastit deflection, due to rotation O(9, and is calculated according to Re ference 10. Penetration (X) of the casb into the concrete structure is obtained t.. the nodified Petry forrula (Reference 8). Deformation ('.) of the af'ected i" pact liniters of the cask during impact is estimated based on the infor+atico Ji ven in Reference 5. Thus, the total distance travelled by the centor o' puity of the cask after the impact is:

^

1 a ! +X+4 (6)

For rr.all drop heights, X and : are nearly zero and thus can be ignored in the tM I / i; The < >uiv41ent force acting on the target structure during impact then equals:

r

~-

@ (7)

C Dynemic e't? cts of this i.pplied force acting on the structure are incorporated by using a dvaamic load factor of two for all shear calculations exceot for edge ima;t on slabs with small drop heights, This is the maximum value for a rectangJIar oulse loading which is conservative for an elastic rescom e (Reference 18). Once the force is obta9ed, the slab ond beam are checked for punching shear and end shear.

4 for Sippl,t h ppor_ted Beams l

ror a simple beam with concentrated load acting on its mid-span, the maximum  !

resistance (R,) of the beam and the deflection (6,) of the beam when its loading ,

reaches the maximum resistance can be determined easily. Using the ductility ratio (u) as recorrended in Reference 12, ti.) mar.imum deflection (6,) which the ,

beam can tolerate equals 96,. The energy which can be absorbed by the beam equals the area under the resistance deflection curve.

Thus, E4 = Area of ABCD (see figure 3 5)

As shown in figure 3 5, the work done by the existing load on the structure due to elastic deflection 6 9and resistance Rg is also taken into consideration, i P

Once E g is obtained, the remaining analysis procedure is exactly the same as described in the previous subsection.

3.4.2.2 Edge ___ Imp _act for Slabs The critical item for cask impact at the edge of a slab adjacent to supporting bears or walls is the shear failure of the slab. Two possible situations are

onsidered:

).arge Height of Cask Drop t For a cask drop from a large height at the edge of a slab, adjacent to a beam er wall sucport, the deformation and penetration of the cask are much larger when compared wit'1 the deflection of the slab due to cask impact. In this case, the deflection of the slab under consideration is conservatively neglected in the analysis. Thus, the force during impact equals fa I

• i~c (7) 1 where:

• d = deformation of the cask plus the penetration of the cask into r.oncrete slob ( f .e. , A 4 X) 3 ,

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The equivalent static force is obtained by applying a conservative dynamic load factor of two. The slab is then checked for punching shear or the shear- .

friction force at the slab support edge. The allowable punching shear for [

concrete is determined by the method given in Reference 13. The foregoing  !

procedure applies to the postulated drop locations at the edge of the spent fuel storage pool at elevation 988'-11" and at the bottom of the equipment hatch at elevation 935'-0" (see figures 3-2 and 3-3).

Smell Heigh _t of Cask Drol >

j for a very small height of drop the deformation of the cask and the penetration i of the cask into concrete slab 17, extremely small. Thus, the effects cannot be determined by the method above. In this case, the longitudinal reinforcement (bottom reinforcing only) passing through the section at the edge of a slab support is used to calculate the shear friction force (S.F f.) according to I

Reference 14 Due to the deformation of the slab relative to the support, a -

1 finite length (L) of the longitudinal reinforcement will undergo tensile strain.

l The length of reinforcement undergoing tensile strain is conservatively assumed t^ be equal to the depth of the slab under consideration, as shown in figure 3 6.

By limiting the maximum allowable strain in the reinforcement to two percent, t

the cermissible elongation of the reinforcing bars in length (L) is obtained.

This e'or.gation is used to calculate a vertical displacement (6 ).

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lhe ecergy balance equation for this case Can be written as:

W (h n + y6)= (Sf.f.)6 y +E, i where: '

l W = weight of the cask hm= maximum permissible height of drop for 2.0% strain in -

reinforcerent E

3 = strain energy stored in length (t.) of the reinforcing bars 10 -

f rom the above equation, 'br maximum drop height (b,) is de W ned. W factor of safety is obtained by dividing h, by the controlled lif t Feight (h,)

above the target slab. The foregoing procedure is used to evaluate edge impact adjacent to support structures along the path of the cask over the operating floor at elevation 1027'-8".

3.5 RESU,LTS NR CONCLUSIONS The results of this investigation for the NSf 4 cask are tabulated in Tables 3-1 and 3-2. Results include a tabulation of key parameters used in the investi-gation and also the resultant factor of safety for each location. Two types of factor of safety are evaluated, depending on which mode of behavior was most critical, for areas where energy considerations dictate the critical condition, the factor of safety was determined by dividing the maxirum permissible cask drop beight (h ) by the controlled task lif t height (hg ). For areas where shear was the critical condition, the factor of saf ety was determined by dividing the maxinum permissible shear (V u

) by the shear stress calculated for the particular drop height under investigation (V), Drop location Numbers 1, 2, 3 and 4 were analyled by the energy balance method. from the results pre-sented in Tables 3-1 and 3 2. the maximun lif t height of six inches above th?

operating floor is safe. The minimun factor of safety for this area of investigation was 1.23 for edae impact, ih' analyses also demonstrate that the spent fuel pool floor slab is capable (if stevivi .; both the vertical drop at the edge of the slab and an assumed her1.:n: 3' 17act at tFe middle of the slab from six inches above the operating float e b u.1on 1027'-B". The minimum factor of safety for the spent fuel pool On a'as fcano to be 1.5 for location No. 5 at the cask setdown pad. For impact en tre setW pad at the edge of the pool slab, shear is the critical node of behavior at.d with this large factor of safety cracking should not occur. for impact in the pool center region, drop location No. 6 , the flexural rode of resit tance is employed to abscrb the cask inpact. Although flexural cracking 3 __ _ _ _ _ _ _ _ _ _ _ _ - - - - - - - - - ---

in this area may develop, cricks will not extend through the thickness of the slab, consequently, no leakage is expected to occur. This conclusion is also strengthened by the fact that the mitigating effects of the steel liner plate, the setdown pad the control rod and fuel racks have not been accounted for in the analysis. I l

DroplocationsNo.7and8atgradelevelelevation935'0"(beneaththeequip-rent hatch opening) were analyzed for a drop height of 93'-2" (from 1028'-2") i and 10'-0", respectively. The analysis results show that the 24 inch thick slab below elevation 935'-0" at drop location No. 7 in the equiprent hatch open-ing will not sustain a 93'-2" free drop of the cask. However, administrative contro' can be implemented in this case to essentially preclude the likelihood of a cask drop in this location. These controit are discussed in Section 6.0.

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TAult 3-1 RESULTS FOR CENTRAL IMPACT (NFS-4 Spent fuel Cask)

STRltCTURE ITEM Composite Floor Steel Pool Slab Beam Beam Slab W21 x 55 W36 x 160 Cesk Impact location No. 1 No. 3 No. 4 No. 6 i Ultimate Positive 225.2 in.-k/ft 15836 in.-k ?7000 in.-k 24404 in.-k/ft Vonent Moment 1 Mu j Negative ?25.2 in.-k/ft ---- --- 11768 in.-k/ft j Mome r.t Maximun Compressive 0.00563 0.00675 ---- 0.003 Strain in Corcrete ForHinges,'u ITotal

!Deflec- 3.066 in. 6.326 in. --- 3.48 in.

Itien for e t

Oreg fro ~ _ _ _

!Heirht

!*m ---- --- 6.17 in ----

i i

Shear Capacity V 655k 153k 336k 14203k u -

7 ,

Stea- Fo-ce V 309k 139k 334k 496k ,

h.d mur alloweble Dron t eight from r n .3 ray 8.33 in. 30.12 in. 52.5 in. ---

hisoce Ipros:'. J.,)

vasinum Feiont o' jLif t Above Fir. E1. 6 in. 6 in. 6 in. 6 in.

1027'-B" th3 )

Factor of Safety ..

h,/h, 1.39 5.02 8.72 3.95 I

Shear at end of beam for cask drop from height h

• • Factor of safety for pool slab = V u/Y 13 -

1ABLE 3-2 RESULTS FOR EDGE IMPACT (NfS4SpentfuelCask)

STRUCTURE ITEM Floor Pool 24" Equipment 68" Equipment Slab Slab Hatch Slab Hatch Slab Cask Impact location No. 2 No. 5 No. 7 No. 8 Teletration into --- --- 1.45 in. 0.16 in.

Iconcrete X l Maximun Tensile --- ---

, Strain in Rebar c 3 0.02 --

Vertical Displace- ---

jJent at Edge e y 2.18 in. --- ---

201.4 12065 1615 5062 lSheerCapacityV u k k k

• 8067 11831 4019 A:tual Shear V 1arinum Allowable Dro;. ---

.eignt from Energy 7.36 in. --- ---

Beiance Approach (hm)

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• • The slab cannot sustain a 93'-2" height of cask drop.

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4.0 FYl.LVATICN OF CASK HANDLING SYSTIMS The consequences of an fiFS-4 (Appendix A) shipping cask drop accident were evaluated for the plant structures in Section 3.0, Similarly, an evaluation of the crane handling system for Monticello has been conducted to deonstrate the factors of safety existing in the trane system and the degree of single failure protection provided. These evaluations provide inforretion on the vulnerability of the crane to postulated failures and permit identification of inherent design features which enhance safety.

4.1 CPME SYSTEM S_AFETY MARGlt4S The reactor building crane is an indoor, electric powered traveling bridge crane with a single trolley cab or pendant controlled 85 ton main hoist. The crane was constructed by the Crane Manufacturing and Service Corporation to the require-ments for a Class A crane established by the Electrical Overhead Crane Institute Specification 61. A conplete description of the crane handling system is given in Reference 2.

An evaluation of the f actors of safety for the crane mechanical components was conducted to determine the actual safety margins involved with handling of the NFS-4 cask, for comparison, the crane safety factors at the design rated load (85 tons) and for handling the General Electric Company IF-300 (Reference 3) spent fuel shipping cask (70 tons) were also calculated. The results are reperted in Table 4-1 based upon the manufacturers design.

Tne results of this evaluation show that all factors of safety for the hoist corpMents are greater than a factor of 15 when handling the fifS-4 cask and

• arly three tines higher than those obtained using either the IF-300 cask or who lif ting a design rated load. Thus, even though the main load path of the crane hoist system is not redundant. the larger safety factors involved with the handling of the NFS-4 cask make the li6elihood of a main load path failure very rerote.

4.2 CP.ANE SYSTEM FAILURE ANALYSIS This section evaluates the vulnerability of the electrical and mechanical equip-ment of the reactor building crane system to postulated single random failures.

1

A qualitative analys has been conducted to detennine the single randtra failure characteristics of the desian u . i cilure Mode and Ef fect', Analytit, (FMEA) proc e dure. Results for the rechanical components are presented in Appendix B.

In this evaluation, emphasis was placed upon failures not considered in initial design and upon operational errors permitted frun the characteristics of the design, failures resulting from citrere external environment and natural causes were not considered in the evaluation.

4.2.1 Sunnary_ of Results An analysis of the General Electric Company Marspeed 320 Crane Drive System has detemined that there are no single rando") electrical failures, that cannot be compensated for by operator action, which will cause an uncontrolled lowering of the load.

Based upon the analysis reported in Appendix B the f ailure of the following mechanical or structural components of the bridge, hoist, or trolley will result in the loss of crane load path integrity:

1. Main hook (item 1)

2. Hook Swivel (item 2)

3. Hoist Rope (item 3) 4 Equiliter sheave pin (item 6)

6. Drum shaft (item 7)

6. Drum bearing support (item 8)

7. Drum (item 9)

8. Each gear, pinion, and shaft in the drive train (items 10 through 12)

9. Either t;ridga girder (items 22 and 23)

Each of 'tese conponents has been analyzed in Appendix B as indicated by the reference item.

4.2.2 Analytical _ Conclusions Based on the results reported in Section 4.2.1. it is apparent that the crane system is not designed to be single failure proof and is subject to single random failures. Only those rechanical f ailures which result in a loss of load path integrity are considered serious. However, such mechanical failures must 2 -

. 8 be the consequence cf a sudden catastrophic failure of either several composite structural clerents (i.e., wire es) or a single major structural compor,ent such as a large beam or shaft. failures of this type in heavy mechanical equipment (e.g., steel structures) aie highly unlikely and would be precluded in this case by large structural / mechanical safety factors. Based on the results I

from a safety factor investigation reported in Table 4-1, actual crane capability for handling the NTS.4 cask is from 17 to 62 tines greater than required. These increased safety factors provide inherent protection against these mechanical component failures when handling the NFS-4 cask. Turthennore, functional check-out ind periodic maintenance will be employed to ensure crane system mechanical and electrical integrity and to verify crane function for critical load handling operati ons.

4.3 CRANESYSTEMMAINTENANC1 The maintenance and inspection of the reactor building crane is perfonted in accordance with the Monticello Nuclear Generating plant procedures. These pro-cedures will include the following maintenance activities.

1. Lubrication performed on a regular basis as established by the crane manufacturer. Lubrication follows manufacturer's reconmenda-tions as to points of lubrication, frequency, levels of lubricant, and types of lubricant used.

2. A visual inspection of the main hoist rope will be made prior to

- each series of critical lif t operations. All rope inspections will meet the requirenents of ANS! B30.2.0 and OSHA standards as cited in Title 29. CTR 1910.179 (References 15 and 16),

3. Hooks will be inspected prior to critical lif t operations in accordance with the requirements of References 15 and 16, 4 All other crane components including limit switches, brakes, power supplies and control systems will be maintained or inspected in accordancewiththerequirementsof29CFR1910.179(Reference 16).

These maintenance activities will be carried out on a periodic basis as established by References 15 and 16.

3 -

i

TABLE 4 1 REACTOR BUILDING CRANE HOIST SYSTEM MINIMUM FACTORS OF SAFETY

_ CASK HANDLING RATED WITH WITH LOAD IF-300 NFS-4 CRANE _ SYSTEM CCNPONENT J81 (70T) (25T)

Drum 5.6 6.8 19.0 Motor Pinion Shaft 18.5 22.5 62.0 15t Interim Pinion Shaft 7.9 9.6 26.9 2nd Interim Pinion Shaft 5.5 6,7 18.7 Drum Pinion Shaft 7.4 9.0 25.2 Drum Shaft 7.1 8.6 24.1 Motor Pinion 11.0 13.4 37.4 ist interim Pinion 9.0 10.9 30.6 2nd Interim Pinion 6.7 8.1 22.8 Orum Pinion 5.7 6.9 19.4 Motor Gear 10.3 12.5 35.0 1st Interim Gear 8.4 10.2 28.6 2nd Interim Gear 6.7 8.1 22.8 Drum Gear 6.0 7.3 20.4 Cables 6.0 7.3 20.4 Main Hook (Tension in Stem) 6.0 7.3 20.4 Welds (Combined bending and tension; 5.0 6.1 17.0  ;

based on ultimate strength of naterial in welds)

Includes static and dynamic loading l

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e 5.0 E'.M.lVATION Of PLANT SAFETY FE ATURES An evaluation of the effects of a spent fuel shipping cask drop accident for the Monticello plant involves the determination of plant safety features integrity, including:

e Spent fuel Pool Integrity e Spent fuel Integrity e Safety of Vital Equipment and Systems The structural, mechanical systems and nuclear fuel handling equipment at the Monticello plant have been re-analyzed for damage in "esponse to a Nuclear Reoulatory Comission request for information (Reference 1). The results of these evaluations are summarized in this section for the NFS-4 shipping cask.

5.1 INTEGRITY Of SPENT FUEL p00L Based on the results of a conservative analysis of the spent fuel pool floor structure as reported in Section 3.0. Structural failure of the spent fuel pool floor will not occur if the NFS-4 shipping cask were dropped at any of the following locations (see figure 3 2):

A. From an elevation of 1028'-2" with the pool water level at the mininum allowable, in a vertical orientation directly above the cask laydown pad in the south side of the fuel pool.

B. From an elevation of 1028'-2", with the pool water level at the minimum allowable, in a horizontal orientation directly above the center of the fuel pool, for Case A, punching shear of the slab is the dominant structural concern.

Since the results of structural analyses show that the allowable punching shear of the slab is larger than the force acting on the slab during cask impact, significant floor damage or slab cracking is not expected to occur. For Case B, the pool slab will resist the cask impact in bending, flexural cracts can occur as the consequence of formation of yield lines in the slab, but cracks i

l l 51 l

l l

through the entire '1:or cre not predicted. Local defomation and minor loss of integrity of the steel liner r:;, occur for all cases postulated. However,  !

since the integrity of the fuel pool floor structure is not jeopardized, the 4

possibility of significant leakage is precluded.

Any loss of fuel pool water inventory would have to leak through flexural cracks  !

and diffuse through the six foot thick fuel pool slab due to the 16 psi static head at the pool bottom. Therefore, only a minimal loss of water from the pool (which could result from splashing or seepage) must be considered. Any small

! amount of water loss will be limited to the operating floor at elevation 1027'-8" or the level below the fuel pool floor at elevation 962' 6". This would not  ;

impose a flooding problem for any critical plant equipment or plant safety features.

Spent fuel pool makeup systems can provide 100 gpm and standby makeup will provide another 450 gpm from the condensate backwash subsystem, although makeup is not i required for such small losses. Fire system water sources are also available as ,

backup to provide an independent source of water in the event of emergencies.

5.2 PROTECTION OF SPENT FUEL, Spent fuel cask handling procedures require that the shipping cask not be moved t over the reactor vessel or over stored spent fuel in the spent fuel storage pool.

Furthermore, the cask will only be lifted above the spent fuel pool in the designated area above the cask laydown pad on the south end of the fuel pool floor (Figure 5-1). The cask laydown pad is a stainless steel plate 7'-0" by  !

7'-8" and the pad is one inch thick. It is located at the bottom of the fuel pool at elevation 988'-11".

Based upon the current spent fuel inventory, it is not necessary to store spent fuel in the south half of the storage pool. Thus, in order to provide the maxi.  ;

num separation of spent fuel frce cask handling areas, spent fuel will be kept i only in the north half of the fuel pool during cask handling operations over i the pool, as indicated in Figure 5-1.

The inside dimensions of the fuel pool are 40 feet by 26 feet. The 25 ton NFS-4 l

shipping cask is approximately 17'-10" long (with impact limiters) and 16'-10" I

long (without the upper impact limiter). The maximum cross section of the cask l

l 2 -

E,. -_ . __ __ __ ._ _ . . . _ _ . - - -_ . - - ~ . - _ -. _ - . . _ . . _ , - _ _ - - . - .

is approximately 60 ' ches in diarcter. In the highly unlikely event that the fuel cask might be deflecten vi  ;< over into the spent fuel pool, the cask site is net adequate to reach and damage those spent fuel elenents which might i

be stored in the north one-half of the spent fuel pool. Furthermore, only a

limited chance exists for a cask tipover situation to occur since the cask 1 would have to be dropped directly on the edge of the fuel pool. If it is 1 postulated that the cask were dropped at this position, the area of influence l resulting from tipover in the south one-half of the fuel pool is shown in  !

l Figure 5-1. The 214 inch cask length would not reach the center point of the  ;

fuel pool assuming that a rotation into a horizontal position were possible '

as indicated in figure 5-1. Moreover, if the cask were dropped on the pool edge, its impact would cause the pool edge to spall and force the cask into the fuel pool in a nearly vertical attitude. Thus, a horizontal cask drop is considered '

highly unlikely due to the site and mass of the cask. '

I A stability analysis has been conducted to determine the need for lateral or  ;

seismic restraints to ensure that a free standing NFS-4 cask will not impose a potential threat to fuel shipping operations due to cask tipover. The analysis nethod is given below. Taking the weight of the NFS-4 cask (W) and the following force diagram:

a f I i (See page 5-4)

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F) = f g

• 50,000 lb f Using figures 5-2 and 5 3, the north-south acceleration of the reactor building is 0.159 and the east-west acceleration is 0.169 at elevation 1027'-8". Addi-tionally, a vertic.a1 acceleration of 0.04g is added to be compatible with the postulated seismic conditions. Using these accelerations, the forces acting on a  ;

cask are:

F = (0.049 ) = E,000 lb f 2

r 3= {- (0.169) = 8,000 lb f l

l Applying these forces at the center of gravity yields the resultant force R and the associated angle a using a conservative static approach:

54-

1 . ,

1. .

l .

Thus, tan'I a a h .107 a= 9.46' R = 48,700 lb f

Taking the moments about the lower right corner of the cask gives (clockwise is negative):

6 (F)-F2 ) x 25" - F3 x 105" = 1.2 x 10 - 8.4 x 10 5 a 3.6 x 105 in-lb f 5

The cask has a net positive n'oment of 3.6 x 10 in-lb which implies that no overturning moment exists during the design basis earthquake (DBE). The reference Monticello DBE has a 0.069 ground acceleration as established by the plant FSAR. As a verification on this calculation, Reference 5 states in Section 3.1.6.1.5, " Bottom Corner Impact 30 foot Free Fall", that the critical angle of tilt such that the line of force passes from the C.G. to the bottom corner of the cask is 13,5'. Since the resultant angle a is only 9.46', the force R is not applied at or beyond the bottom corner of the cask. Thus, cask stability is assured without restraints. The analysis is considered conserva-tive since the acceleration forces are less at the fuel pool elevation 988'-11" and since the actual loadings are not constant but oscillatory in behavior.

Baseo on this analysis, it may be concluded that there is no need for lateral or selsnic restraints in the fuel pool to protect the spent fuel, or at the decontamination pad to protect the reactor building operating floor from a

.(si ticover accident.

5.3 PROTECTION Of VITAL EOUIPMENT Detailed structural impact analyses for the reactor building operating floor, fcr tre areas beneath the equipment hatch, the spent fuel pool, and the cask decontamination pad have been completed. Based on these analyses, it may be concluded that the building structures are adequate to withstand the impact from a postulated cask drop accident. Therefore, no damage to vital plant components or systems and to non-vital components or systems will occur.

When the NFS-4 spent fuel shipping cask is unloaded from the cask trailer in the cask handling area at elevation 935'-0" (grade level), the cask will not be 5 -

i raised any more than 10 feet above the floor area to clear the trailer. Before hoisting the cask above this elevation, the suspended cask will be moved to a position 7'-0" west ,nd 8'-0" north of the hater, opening (see Section 6.1).

1 This will locate the cask directly above the 36 inch diagonal wall below the floor at elevation 935'.0". This is the most structurally sound section of the equipnent hatch floor area as determined by the structural analyses in Section 3.0. Once this position is reached, the suspended cask will be hoisted above the 10 foot level to the operating floor. This procedure will prevent the cask from being hoisted above a level for which damage could occur to the structure and the containment vessel below elevation 935'-0". The 36 inch-diagonal wall (figure 3-3) will mitigate the effects of a cask drop for hcights greater than 10 feet in the specified hoisting position.

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6.0 ADMINISlRAli"E r~ 'e0' 0F CASK HANDLING OPERATIONS Administrative control of plant activities during cask receiving, loading and shipping operations will be established in accordance with specific orarating procedures. Special safety precautions provide added protection and assurance that a cask drop accident would be precluded. Planned activities to mitigate such accidents include a fixed and strictly enforced travel path for cask move-ment, pre-lift checkout of the reactor building crane system and detailed pro-cedures for cask receiving, loading and shipping operations, i

6.1 RESTRICTED CASK MOVEMENT PATH During cask handling operations, the spent fuel shipping cask will be received at the site en a transport trailer via the reactor building air lock access at elevatien 935'-0". The shipping cask will be removed from the trailer by the reactor Duilding crane and hoisted to the operating floor at elevation 1027'-8" where it wi'l be roved to a laydown area and then lowered to the bottom of the fuel pee' at elevation 988'-11". Following spent fuel transfer into the cask, th- cask will be renoved and loaded onto the transport trailer for offsite

-4  : This sequence is shown visually in Figure 6-1.

no

• icted travel path is also shown in Figure 6-1. The travel path is c~ rati c 4 with the crane operating envelope shown in Figure 4-1 and the rasults

~ 4tructural analyses reported in Section 3.0. The cask will be hoisted in ?csition A, to an elevation of 1028'-2", or less, prior to horizontal travel acr 55 tha operating floor. The cask will be moved from Position A to B at alesat on 1028'-2" to avoid movement across the hatch opening or over the torus 4

stracture below elevation 935'-0". Movement of the cask along sections A-B, l

B-C. C F, D-E and F-G will ensure that the cask is restricted to those locations caceble o' withstanding the structural impact.

l l The travel path along Section B-C was chosen near the center of the 22'-9" scan between column lines 7.9 and 8.9 in Figure 6-1. '-avel path C-D was chosen at the midspan between column lines P and R to maximize structural cap-ability. Travel path D-E is near the midspari between column lines 6.9 and 7.9.

The cask travel path is not in direct correspondence with the midspan since D-E 1

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must intersect with we c:sk setdown pad in the fuel pool. Point E, directly above the cask setdown pad, corresponds to a postulated fuel pool drop case from six inches above the operating floor, The cask return travel path is E-0, 0-F and F-G in Figure 6-1. Similar care has been given to this travel path to restrict the cask movement between column lines P and R or 6,1 and 6.9 to maximize the structural capability and safety factor margin. The total cask movement would then be A-B-C-D-F-G for receipt and inspection, G-F-D-E for loading in the spent fuel pool, E-D-F-G after loading for decontamination, and G-F-D-C-B-A for cask return to t'le transport trailer and shipment.

To ensure movement of the shipping cask along the designated path, fioor markings will be made with a bright color as indicated in Figure 6-1 to guide the crane operator and plant personnel during cask handling.

6.2 CRANE SYSTEM PRE-LIFT CHECKOUT A review of crane operating requirements at the Monticello site has been completed in conjunction with plans for spent fuel cask handling. From this review, a detailed procedure will be developed to provide instructions for a functional checkout of the reactor building crane to be perfcrmed prior to any series of critical hoisting operations. The activities disCJssed in this section provide a sumary description of the detailed pre-lift checkout procedure to be implemented at Monticello.

6.2.1 P_ersonnel, Safety, Equipment Protection and Operability The pre-lift checkout of the Reactor Building crane will be perfonned by a person or persons who are familiar with trane system cperation and plant procedures requiring crane function and cask handling. The crane operator will be instructed in cask handling requirements and will be familiar with safety practices adopted by Northern States Power Company. The operator will verify the integrity of the crane system by visual inspection prior to functional sneckout. The crane load shall not be moved over the reactor vessel, the new fuel storage racks or the spent fuel storage racks in the pool during the performance of this checkout.

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6.2.2 Functional Prea.ift Checkout I

The safe operation of the resctor ouilding crane will be verified prior to each series of critical load handling operations in accordance with operating procedures written specifically for reactor building crane pre-lift checkout.

This checkout procedure will call for bridge, trolley, main hoist and auxiliary hoist operation to verify function within prescribed design limits. Brakes, limit switches, interlocks and controls will be checked for both cab and pendant operating modes, 6,3 PROCEDURES FOR RECEl(ING, LOADING AND SHIPPING A review of the planned spent fuel cask handling operations at the Monticello site has been completed. From this review, procedures will be developed to provide the necessary instructions required for safe and effective handling of the NFS-4 cask during conduct of the spent fuel shipping program. The activ-ities described in this section provide a surnary description of the receiving, handling, loading, and shipping procedures that will be implemented.

6.3.1 Receipt of the Spent Fuel Ripping Cask The NFS-4 cask is shipped from the fuel reprocessing plant or fuel storage location on a specially built trailer with cradle mountings for the cask. Upon arrival at the Monticello site, the cask trailer is backed through the railway access air lock and into the cask handling area as illustrated in Figure 6-1.

Secondary containment shall be maintained at all times during the fuel handling by either the inner or outer doors on the railway access air lock of the resttar building.

The reactor building crane 85 ton block is lowered into place above the stored lifting yoke on the trailer and the yoke is secured to the crane and cask. As the cast It lif ted, th' task is pivoted about the rotation trunnions to the vert i cal position shown in figure 6-1. The cask will be raised and lowered directly above the trailer to verify the function of the lwded crane. The operator will not lif t the cask more than 10'-0" above the access level at elevation 935'-0" during this verification. The cask will be lifted approxi-mately 93 feet up through the equipment hatch, directly over Position A, and moved horizontally across the operating floor along the predetermined path de-scribed in Section 6.1 and shown on Figure 6-1. The cask is moved to the 3 -

decontamination pad ard sat down in a vertical attitude. During the horizontal movement, the cask shall not be liited more than six inches above the operating floor.

6.3.2 Loading the Spent Fuel Shipping Cas_k The cask is washed down to remove any road dirt and the cask lid bolts are removed. If required, the cask is decontamineted, prior to filling with de-mineralized water and repositioning of the cask lid. The cask is lifted no more than six inches above the decontamination pad and moved horizontally across the operating floor along the path described in Section 6.1 and Figure 6-1 until it is directly above the cask setdown pad located at the bottom of the spent fuel storage pool. The cask is lowered into the pool until it is approximately one foot above the steel setdown pad. The cask is then lowered slowly to the floor of the fuel pool as shown in Figt.re 6-1. The cask lid is removed from the cask and from the fuel pool and the crane trolley is repositioned to allow clear access for the fuel handling bridge to transfer fuel from the stored fuel racks to the open cask.

Handling of the spent fuel within the fuel pool and loading the spent fuel into the '.ask will be performed in accordance with Monticello spent fuel pool handling procedures. A licensed Senior Reactor Operator is directly in charge according to Section 6.1.C.3 of the Technical Specifications. Transfer of spent fuel is accomplished using the refueling platform and associated fuel handling tools.

Two fuel assemblies are moved from their spaces in the pool storage racks to the snipping cask using the spent fuel pool handling tools and the fuel handling bricge. The cask lid is returned to the pool, repositioned on the cask and the cask is lifted from the pool, washed down and the radiation level measured. The cask is moved horizontally across the operating floor to the cask decontaminaion pad. The cask lid is secured and the temperature rise is measured to verify l cooling requirements. After verification that the fuel element heat generation

! rate is acceptable, the cask cavity is pressure-tested. The cask is decontaminated to a level established by Department of Transportation requirements.

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operating floor along the l Spent Fuel Shipp W Shiprent of tN:  ;;;9 tally across the During the horizontal 6.3.3 The cask The cask is raised and moved hcn6.1 to the boveequipment the deck. hatch.

hatch equipment path described in Section lifted more than six inches ain figure 6-1 cask is then The movement, the cask is not rear (bottom) age the is moved over Position A shown than 10 feet above the floor.

opening until the cask is lessil the cask rotation trunnions engvertic cast receiving moved over Loweringthe trailer cask from the unt moved into the of thetransport trailer is a vertical attitude.

cradle saddles. k lif ting cables in l

position is continued as themaintain the 85 ton crane b oc the transport area to ed and the cask diationis secured on survey.

The cask lif ting yoke is disengagt upon conpletion of a ra trailer for offsite shipmen l

6

-___m______ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _

6.3.3 Shiprent of tS:: _ Spent _ Fuel _ _ Shipping Cask _

The cask is raised and raoved hor  :'nally across the operating floor along the path described in Section 6.1 to the equipment hatch. During the horizontal movement, the cask is not lifted more than six inches above the deck. The cask is moved over Position A shown in Figure 6-1 and lowered in the equipment hatch opening until the cask is less than 10 feet above the floor. The cask is then moved over the trailer until the cask rotation trunnions engage the rear (bottom) cradle saddles. Lowering of the cask from the vertical to the horizontal position is continued as the transport trailer is moved into the cask receiving area to maintain the 85 ton crane block lifting cables in a vertical attitude.

The cask lifting yoke is disengaged and the cask is secured on the transport trailer for offsite shipment upon completion of a radiation survey.

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7.0 REFERENCES

1. D. J. Skovholt, letter to L. O. Mayer, Director of Nuclear Support Services, Northern States Power Company. from U. S. Atomic Energy Commission, Directorate of Licensing; Re: License No. DPR-22, February 4,1974,

2. L. O. Mayer, letter to K. R. Goller, Assistant Director for Operating Reactors, U. S. Atomic Energy Conmission, from Northern States Power Company, Director of Nuclear Support Services; Re: License No. OPR-22, October 1, 1974.

3. General _ Electric Company "!F-300 Irradiated Fuel Shipping Cask."

Technical Description report NED0-10864. July,1972.

4. L. O. Mayer, letter to D. L. Ziemann, Chief, Operating Reactors Branch #2, U. S. Nuclear Regulatory Commission, from Northern States Power Company, Manager of Nuclear Support Services; Re: License No. OPR-22, May 30,1975.

S. Safety Analysis Report. " Spent Fuel Shipping Cask Model No. NFS-4."

Nuclear Fuel Services Corporation, West Valley, New York, December 30, 1971 updated September 29, 1972.

6. U. S. Atomic Energy Cannission. " Structural Design Criteria for Evaluation of the Effects of High-Energy Pipe Breaks on Category 1 Structures Outside the Containment." Document (B), Structural Engineering Branch, Directorate of Licensing, June, 1973.

7. U. S. Nuclear Regulatory Commission. Standard Review Plan 3.8.3, "Other Category : Structures." Directorate of Licensing, 1974

8. Topical Report BC-TOP-9. " Design of Structures for Missile Impact,"

Revision 2. Bechtel Power Corporation, San Francisco, California, Septembe r, 1974.

l l 1 -

i l

9. Chelapati , C. V. , Kennedy, D. P. , and Wall,1. B. "Probablistic Assess-ment of Aircraf t Haza6 d for Nuclear Power Plants." First International Conference on Structural Mechanics in Reactor Technology Berlin, September, 1971.

10. Mattock, Alan H. " Rotational Capacity of Hinging Regions in Reinforced Concrete Beams." Proceedings of the International Symposium on Flexural Mechanics of Reinforced Concrete, American Society of Civil Engineers, (PCA Bulletin 0-101),1965,

11. Corely, W. Gene. " Rotational Capacity of Hinging Regions in Reinforced Concrete Beams." Journal of the Structural Division, ASCE , Vol. 92, No STS, Proc. , October,1966.

12. Newmark, N. W. and Haltiwanger, T. W. " Air Force Design Manual."

AFSWC-TOR-62 138. Prepared by the University of Illinois for Air Force Special Weapons Center, Kirtland Air Force Base, N. M., 1962.

13. ACI Manual of Concrete Practice, Part 2, American Concrete Institute,1973.

14 American Concrete !astitute ( ACl). " Building Code Requirements for Re-inforced Concrete." Code ACI-318-71, 1971.

15. American National Standards Institute ( ANSI). " Overhead and Gantry Cranes ,"

USA Standard B30.2.0-1967, by ASME ,1430 Broadway , New York, N.Y. ,10018.

16. Title 29, Code of Federal Rcgulations, Chapter XVil- Occupational Safety and Hedith Administration (OSHA), Part 1910, Section 1910.179 -

Overhead and Gantry Cranes.

17. American National Standards Institute (ANSI). " Hoisting, Rigging and Transportation of items for Nuclear Power Plants," Draft ANSI Standard N45.2.15, Draf t No. 2, Rev. O, May ,1974,

18. Horns, Hansen, Holley, Biggs, et al, " Structural Design for Dynamic Loads,"

McGraw-Hill Book Co., 1959.

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I APPENDIX A '

I r

i' DESCRIPTION OF THE NFS-4 i

SPENT FUEL SHIPPING CASK  !

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DESCRIPTION OF THr urS-4 SPENT FUEL SHIPPING CASK The NFS-4 spent fuel shipping cask will be used by Northern States Power to ship irradiated fuel from the Monticello Nuclear Generating Plant site. This cask is manufactured by Nuclear Fuel Services, Incorporated, of West Valley, New York. The cask has a capacity for two BWR fuel assemblies and its loaded weight is less than 50,000 pounds. The NFS-4 shipping cask has been licensed for use under Title 10, Code of Federal Regulations, Part 71. The bases for this license are docurented in Reference 5.

The overall dimensions of the cask are 214 inches in length and a maximum of 50 inches in diameter. The cask is shipped on a transport trailer inside an expanded netal enclosure to provide personnel protection from the heated cask surface. In transit, the cask is supported horizontally by cradles mounted to a truss on the transport trailer. The trailer is towed by a tractor. The tractor-trailer assembly weighs approximately 73,000 pounds and the combined length is less than 60-feet. This size is in accordance with rules imposed by the Department of Transportation for the Shipment of Radioactive Materials.

The cask is cooled by natural air convection and thermal radiation and requires no auxiliary cooling system. The fuel in the cask is cooled by natural convection flow of the cavity water. The cask cavity is fitted with a removable basket which holds two rectangular fuel canisters. The canisters are perforated to allow circulation of the cask cavity water around the fuel assemblies. The heat fron the cavity water is conducted through the cask, gamma shielding and neutron shielding to the outside cask surface which is cooled by air. This ccnfiguration is capable of dissipating 42,000 Btu /hr.

The cask t.ody contains a 6-5/8 inch lead gamma shield in an annular geometry between two cylindrical stainless steel shells. These shells are welded to a solid 8-inch stainless steel bottom section, which serves as a lower gama shield, and to a solid stainless steel flange section which, along with the cask lid, forms the upper cask closure and the upper gamma shield. The cask lid is a flanged frustrum of a cone with a maximum thickness of 7-1/2 inches.

The flanged section of the lid is 2 inches thick and 26-1/2 inches in diameter.

- A _ . _ . _

The lid flange is counter-bored with six holes for lid closure bolts and the top side of the lid is counter-bored with four blind-threaded holes for the attachment of the upper r,ask impact limiter. The under side of the lid flange is grooved for the two 0-ring seals. The cask walls are surrounded by a compartmented annular shield tank between the cask ends. The tank contains a borated water neutron absorber solution for neutron shielding.

The cask is protected from structural impact damage by two stainless steel clad, balsa wood filled, impact limiters on the top and bottom of the cask. The top impact limiter is removed by loosening the four bolts and sliding it away from the cask lid along tracks prior to lifting the cask off the transport trailer. The lower impact limiter is an integral part of the cask base. The outside diameter of the lower impact limiter is 50 inches which provides a stable base for the cask when it is to be positioned in the vertical attitude. ,

The cask is fitted with two 8-5/8 inch diameter by 3-inch long flanged trunnions on the outside of the upper cask impact limiter. The cask is lifted by a handling yoke which attaches to these upper lifting trunnions. The lower section of the cask is fitted with two 6-5/8 inch diameter by 3-inch long flanged rotation trunnions which are offset from the cask centerline to facilitate cask rotation. This offset location ensures that the cask will rotate into a horizontal position when lowered on the trailer. On the transport trailer, the cask is supported by the upper lifting trunnions and by the lower rotation trunnions. The trunnions are cradled and locked onto front and rear trailer supports.

The cask is fitted with a leak detection valve for sampling the area between j the two 0-ring seals on the lid. The cask body is fitted with a cavity vent

! valve, two cavity drain valves, and a pressure relief rupture disc which is backed up by a pressure relief valve.

The cask is designed to withstand, with only limited release of radioactive materials, a 30-foot free drop onto an unyielding surface, a subsequent 40-inch drop onto a six-inch diameter pin followed by 30 minutes in a 1475* fire and eight hours immersion in 3-feet of water. These requirements form the accident l design bases for the cask in accordance with Title 10, Code of Federal Regula-tions, Part 71, Section 71.36 and Appendix B. The cask is classified as a j - A -

Category A item according to ANS! 'JE.2,15. Hoisting, Rigging and Transporting of items for Nuclear Power Plants (Reference 17). The cask is classed in Transportation Group III according to the Code of Federal Regulations, Title 49 Parts 171-178 Department of Transportation, Regulations for Transportation of Radioactive Materials as amended.

A summary of the design characteristics of the NFS-4 spent fuel shipping cask is given in Table A-1. Figure A-1 shows the design configuration of the NFS-4 fuel shipping cask.

l

- A l l

I __

TABLE A-1 ,

DESIGN CHARACTERISTICS OF NFS-4 SPENT FUEL SHIPPING CASK (Per Ref 5)

Fuel Capaci_ty: 2 BWR Fuel Bundles Cooling Capacity: 42,000 Btu /hr Loaded Weigh 1:

Cask and Fuel 49,464 lb Tractor, Trailer and Loaded Cask 73,000 lb Cask Body:

-Length (including impact limiters) 214 inches Diameter (maximum) 50 inches Cask Cavity:_

Length 178 inches Diameter 13-1/2 inches Cask Materials: Stainless Steel Shielding Materials: Lead and Borated Water Impact Limiting Materials: Balsa Wood in Stainless Steel Cask Cwity Design Rating: 170 psig @ 3520F 0

Cask External Design Temperature: -40 F to 130 F Cask Lifting Devices: Yoke and two 8-5/8 inch diameter by 3-inch long flanged trunnions Safety Requirements:

Regulatory Criteria 10 CFR 71 Impact Design 30 ft drop to unyielding surface.

40 in, drop on 6" dia, pin Extreme Environment 30 min in 14750 fi re 8 hr. in 3 fee

• water

- A ,

LOWER IMPACT LIMITER G AMMA SHIELD CASK CAVITY-CASK LID w 4J 1

fk,l.;((j.

\

/N kl[

j- NEUTRON SHIELD TANK hh.h27 ML A' SURGE TANK 0 ' gs4,. m.

LIF TING TRUNNION UPPER IMPACT LIMN ER LID IMPACT LIMITER 4

NFS-4 SPENT FUEL SHIPPING CASK Figure A-1

APPENDIX B FAILURE MODE AND EFFECTS ANALYSIS FOR THE 85 TON REACTOR BUILDlHG BRIDGE CRANE AT THE MONTICELLO HUCLEAR GENERATING PLANT

e f A I I t? ft E E3 O D F; A N D E 5' f f. ( T ** A N A I T $8 5 ",

Tage 1 ef 5

________q._.__________ - _ _ _

4 nUCLCF.R ' Fregram W15?f f rf 10 fFA*if EYalt RIIM e_r-st:n.1 level Df ur== . ' p.m ,,,,,, , , , ,

systes R[, ACTOR 6;!!Dif.G CRANE Sf57EM gg g g j gggy,ggg gg g gygg ggpp g 7gg gg ,;,g3 , CORPCE Alloa , , , , , , , , ,, g g s,3 t , ,q 0-514-8; D-514-81; D-514-M3 Pev I- D-514-Hl); D-514 T; g 4 ,,,_

g .g g pg7g s e-a.istem D-Set-T1. Rev 1. Feeleved 4. k;[Q Dete tg.T g 4' ;s? t!C*L (MC*it4TS ED iP "* appre e4 W M T d Dete % g 9

Method Inherest Errect pe.oree go. sese Tellure cause Symptome and Local Effeets of 6

• t 14 @a w Mose Imeludleg Dependent F=11erce Deteetton Provision Oystem Other Effeets i

1. watn ute 1 Mote 2 Uncontrolled drop of lead. Self - lacreased safety fatter load will N0ch annunClating m:th less than rat *d drop.

load rabes protablitty of t*.ts type of f all.

wre estewely small.

2. rmok Note 1 Note 2 Uncontrolled drop of load. Self - Increesad safety fattor Load util Swivel annunciating =1th less team rat *d drop.

load metes protability of this type of fall-are entremely small.

3.  ?*ain hote 1 Note 2 Uncontrolled drop of load. Self - Increased safety facted load will Hotst annunciating with less than rated  ; drop.

Rope load rebes prebability j cf this type of fa11 are entremly sea 11.

4. w a tr Mote I Note 2 Load displaced vertically by 2.0 Self - Increased safety facto + Ur.able ts i fw.hes . Bottom block frame will annunciating with less t*an rated coerste Nist load asaes probability creae.

60tt n Block catch sheaves.

of this type of fail-5" eave pins are estremely small.

I

1. All postulated fattures are considered to be catastrophic (i.e., rupture, brittle failure, etc.)

2. The casses of these postulated fattures include inherent flaws. overstressing, fatigue, etc.

s -

F A t a U R E MO D E A N D E F F EC TS A N A L Y SIS Fee. 2 ef 5 gegCttas ' Freer DOfflCitt0 (v4C Evatt:ATian reacttenen te. 1 Dt.ar.=

1 g , ,, ,,,

Syste. REACTOR BUILDIC CRAnt 575ftM N IK y,,,,,,, ,, y ((( (gu,J.

5.b-Syst.. CRANE LGAD PATM N IWEJfACT*AIE M SFWVIE CNTilm lutkIfeG5 D*** Y' D- C- I D-514-4; Fe.Sewee 4.y*Q D.te gg_p(

,,.,,...t nm.,= amars m n .D.-514-81;

.e i.

D-514-40. Ihre I; D-514-M13; D-514-I;

,,,, ., y _ g,, m ,y

+

@ t

... . ,. ..r.

c.... sr..

i.et ..e

. ..e m . arr ..

De ,e.. s ... ....

"'"," c.';';;;Ue  %,,.;;' **- ;"  ;

s. s, ,,..,.... _ .s, , ,,,..s.

i

5. Main note 1 note 2 toad displer'd. vertically y 1.125 Self - Increased safety factec Unable to Holst inches. Te,- blott framme s11s catch esausatisting with less them rated jogarete Top Block sheeve pies. leed ashes prwhability trone. .

5 heave of this type of fall-Pips are estrumely semil.

6. teualtzer note 1 mete 2 Urmatrolled W of lead. Self - Increased safety factee Loed will 5 heave ammunriettag with less then rated drop.

Fin loed metes probeb111ty of this type of fall-are entremely small.

F. Mata note 1 note 2 tentantrolled drop of lead. Self - Intressed safety factee Lead will Drum esausacleting with less than rated erup.

Shaft load antes prohobility of this type of fail-are entrymely samtl.

8. Pain Note I Mote 2 Uncontrolled drop of leed. Self - Increased safety facts tend will Drus annunciating with less then rated drop.

Bearing lead makes prebebility 1 5.apport of this type of fall- i are estrumely small. l

9. Main Note 1 Note ? Uuncontrolled drop of lead. Self - Increased safety fact:n toed will Drus asummaciating with less then rated ersy.

Ioad es6es prwhebility of this type of fall- ,

are estremely sam 13 l 10, Main hate 1 note 2 lhicentrolled drop of lead. Self . Increased safety factog toed will Hofst ennsettating with less then rated drop.

Notar lead amnes prehebility Pfeion of this type of fall-  !

eure estresurly sell. 6 4

I

1

/

s

• 1 F A l s t.' R E N O D E A N D E F F EC T S A N A i Y SIS rene 3 of 5 g nycttaa Progree S T tttL2 Cette tystusTi ?

reette.e1 tevet etsgram

& SCSY3CCS peport so. trig.ro_a"S.ns systee REACT 02 EJ:tCI';G CaANE $y3f[w .

Frepued By f d hc.M l sub-syates (2AJE t0A') PATH (PNE MA%TACite!NG ke3 5tpyICE CfSFORATION DPA i1%55 e,te sir.hf re . &

"M*IW C7C' D-514-8; C-514-81; D-514-H3 Ree 1; *)-514-H13; 0 514-T; h'8'"'8 A $![A D'*

• M- 4-1 f Epireent I

D-514-71. Rev 1. Approced 6 M & M ete A)-6-76 5

I Inherest Effect Reserne Pethod So.I sene Fellure Cause Symptees eed Local Effecte of Co*Feaseting tiros sad Mode Including Dependent Failures Detection Freelsten F7 st*w other Effects 11 Fa M hete I hote Z Uncontrolled drop of load. Self - Increased sa'ety factor Load will with less than rated jdrop.

l.

Moist annunciating Icad ubes pret,abt11ty

• ctor of t%15 type of fall-Shaft wre entremly seal).

Uncontrolled drop of load. Self - Increased safety factor. Lead =111

12. Nafn hote 1 Mote 2

! annunciating with less than rated  : drop.

Hetst load ea6es pretso111ty Mctor of t%fs ty;* ef fatl-Gear vre entrecely sult.

Note 2 Uncontrolled drop of Iced. Self - Increased safety factof Load will

13. "atn hote .

annunciating with less then rated drop.

++cfst load ma6es probability First of this type of fail.

Irte-fe vre entrecely smell.

Pfnton Uncontrolled drop of load. Self - Increased safety facto, Loed =111 14 N!1 hote 1 l Acte 2 with less than rated drop.

annunctating

-Ofst load siabes prcbatt11ty First of tels type of f all-Interis ure entre ely smil.

5'a't Mete 2 bacontrolled drop of load. Self - Increas=d safety factc* Load will

15. Patn Note 1 annunciating with less than rated drop.

Hofst load rates probability First of this type of fail-N erim are entreeely smil.

Gear Mote 2 Uncontrolled drop of load. Self - Incressad safety factoe Load will

16. 8 Patn hote 1 annunciating with less tFan rated drop.

Paist load metes probability Second of thts type of fa11-Inter 1m are entremely small, Pinton l

e e A I e t' H f: st o I) 1 A 14 Is t. } F E C T S A f6 A 8 % % t 5 F.g+ a er 5 etc.ette .1 te .i Diser e

,#7 E" #*****

Y SC M CS Perort so. L5 g a.w.cc.0750- M Systee Prn*;M ETt I% CoANE SYST[q

, CORPCRATIO!! p,,,,,,, ,, g , ,

Sub-System CRA'J tcAS PATH CPA%E MANJTACTL1tIhG AND SERVICE C09PORAT!(71 DRAW!%GS D-514-8; D-518-81; D-514-N3. Rev 1: D-514-H13; D-514-T; 3*** \%' h( '"' G

n. ,-e.t n-Icx C--S o-s a-'i. a~ i-

"~' '(sde=~ioer a,,e.ee eu ww.t. -,s o

No. Name Tailure Cause Symptoms and Local Effecto Imeluding Dependest Falluree C re s t g Mode Detection Freelsios SF at** Cther Effects  !

17 I Pain Note 1 Note 2 Uv ontrolled drop of load. Self - Increased safety factoc Load =111 Noist annunciating with less than rated drep.

load ma6es prebaM11ty Secord of this type of fail-laterim are entreawly small.

Shaft

18. *a17 hete 1 Note 2 Uncontrolled c*op of lead. Sel* - Increased safety factof toad will

-cist annunciating with less than rated drop.

Second load makes prebability Ieterte of 1915 'ype of f ail-Bearing are ?stre ely snell.

19. Main Note 1 Mete 2 Uncontrolled drop of load. Self - .

Increased safety facto + Load will dist annunciating with less ttas rated drop.

tew 5 peed load riehes probability Pinton of this type of f all-wre entremely srall.

20. Main hote 1 Note 2 Uncontrolled drop of load. Self - Inreased safety factov Load util hoist ennunctattrig with less than rated drop.

to. Speed load eates prot,atility of this type of fail-5* aft are estremely small.

21 *atn Note 1 Note 2 Uncontrolled drop of load. Self - Increased safety facte* Load wili

• oist aamunciating with less than rated drep.

tc., Speed lead sa6es predet:111ty of this type of f ail-Gea r wre entre-ely smalt.

22. '

Pain hote 1 Note 2 Uncontrolled drop of lead. Self - Increased safety factoi'Loed will Girt annunciating with less than rated ' drop.

load makes prr,bability I of this type of fall-ure extremely small.

+ ,

b

i O.

7 i

F A l l. U R E M O D E A N D E F F E C T S A N A 1. Y S I S ,

^~

4 Fase 5 ef 5 AgCt(M ' Fregree liMllO CW'E EVALU87IC'8 .

Funettonal level Dt. gree Systeo REACTCR SUILDI C C#ANE SYSTEM E Prep.ree my { t I (Ji,M m Sub-Systes CRAME LCAD FATH CRANE MesufAGall!NG AND SERVICE CORPORATION DRAWING 5 D*S' Peetwee I\~1**!'l(

L 4:k _'Date8' - C 1,.3 3g-D-514-8; D-514-81; 0-514-H3, Rev 1; D-514-M13; D-514-T; '

Equipment FECP/JICAL CAPFONENTS 0-5 % T1. Rev 1. ,,,,,,,, yc_. r _.ee, to,g, g, O

" '"

• 2 * "- ' "r~' a- ~

i se. ...e r.li.re c...e S,.,te.e ..e u 1 ser-t. c Mode incl.dtag Depeseemt F.11=ree Det etten F

• Other rette Uncontrolled drop of Ioed. Self - Increased saf-ty fector Load w111

23. Main hote 1 Note 2 with less than rated

• drep.

esmunciating G1rt load sekes prebet>1itty II of this type of fall- .-

are entremely ses11 . .;

i Lead will be displaced vertically by Self - Retainers on trolley Unable to 24 Trolley hote 1 hate 2 . perate 0.5 Inch. annunciating truck prevent drop.

weel crane.

Asle t

Load will be displaced vertically by Self - Retainers on bridge Uneble to

25. Bridge note 1 Note 2 truck prevent drop, operate 0.5 Inch. annunciating wheci crene.

Asle 4

2

.i 4

1 i

. . , , -. - _ - _ - . . . . , . ~ , _ , _ - - _ . - , . _ _ . . - .

q ,- . N R C D lf.1 h JTION l'OR PART tiO DOCKE1 MA 'l AL O LMPORARY FORM)

CONTilOL NO: J12 F I L E : __. -- _._

F ROM: NSP DATE OF DOC DATE REC'D l.Tl1 TWX RPT OTHER Minneapolis, Minn. 55401 g L D. May1r _ _ _ _ . _ _ . _

TO: ORIG =CC _ _OTHER.__ . _SFNT ttRC POR XI__ _

Hro ,. Stello 1 signed 39 SENT LOCAL PDR ..___X.I

~

CLASS UNCLASS PROPINFO INPUT NO CYS REC'D DOCKET NO:

XXX 40 50-263 DESCRIPTION: Ltr trans the followingt ENCLOSURES: "An Analysis & Safety Evaluation of Spent Fuel Shipp6ng Cask Handling at Monticello Plant"....

(40 cys enc 1 ree'd) hh PLANT N AME: Monteicello Plant

. _ _ _ . . ._ _~

Q[(R gg SATETY FOR ACTION /INFC RMATION ti;y r r DHL 1-26-76 ASSIGrED AD ASSIC1'cp Lt: AN:!' c 11EP UKANCll ClllEP k DMN _

PROJECT MM:AGER . _ , _

NROJECT 11AllAGER,.kM.@,M LIC ASST. , _ _ _

W/ ACMS d lC. ASST. M M ( _ U//fCYSACRS INTERNAL DISTRIDUTION CG PIITC SYSTEMS SAFETY PL \NT SYSTE!!S~ SITE SAFEYY

• ENVtMO ANALY"1S WGC PD.t HEREMMI TEDESCO DEttr0N MULLEit 96 ELD SCllROEDER LENAROYA GOSSICP/STAFT LAINAS ENVinD TECH. CITE AllALYS2 S th&E (2) ENGINEERANC IPPOLITO ERNST Vf ID'ER HIPC MACCN;Y bALLARD El'Nrl!

CAEE );NIGllT OPER A?ltC RE ACTORS SPANGLER J. COLLINS PROJECT !!ANAGEMENT S7P.WEIL STEL!h lasEGER BOYU P/.WL1 CK1 S I,T,C, TE CH . ,

P. COLL 11;S OPERATtNC TECif. CAMMILL ATM ll0USTON REACTOR S AFETX idHSENilUT STEPP SillT'ZMAN FETERSON ROSS ' 411 A0  !!Ul}!aN RUTLERG MELTZ t'0VAK MAER llELTE!!ES ROSZTOCZY $ CITENCER M 1 SC) L ANEOUS~

CilECT. dRIMES EK[ yin AL OlSTRIBUTELN __

l / LOCAL paa Minneapolis, Minn. N/ir10NAL L/ sit '!/ CYS IIRUOZAVEN NA'1. L/.1; jdr1C REC 1011 V-1LE-(WALNUT GREEK) ULN1 RSO" (ORNL) dis 1C LA PD't ASLB CONSUllrAfriS 4 %,

.