ML17207A460

Amend to Page 5.4 of Tech Specs for Design Features.Safety Evaluation Summaries for Spent Fuel Storage Rack,New Fuel Storage & Fuel Insp Elevator,Upender & Fuel Transfer Tube Encl

ML17207A460
Person / Time
Site:	Saint Lucie
Issue date:	10/04/1979
From:	FLORIDA POWER & LIGHT CO.
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ML17207A459	List: ML17207A458 ML17207A460
References
NUDOCS 7910120249
Download: ML17207A460 (18)

Text

DESIGN FEATURES 5.2.1.2 SHIELD BUILDING

a. Minimum annular space = 4 feet.

Annulus nominal volume = 543,000 cubic eet.

C. Nominal outside height (measured =rom top of foundation base to the top of the dome) = 230.5 eet.

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d. Nominal inside diame er = 148 feet.

e. Cylinder wall minimum thickness = 3 feet.

Dome minimum thickness = 2.5 feet.

Dome inside radius = 112 feet.

DESIGN PRESSURE .AND TEHPERATURE 5.2.2 The containment vessel is designed and shall be maintained for a maximum internal pressure of 44 psig and a temperature .of Z64'F.

I P ENETRAT ON 5 5.2.3 Penetrations through the containment structure are designed and shall be maintained in accordance with he original design pro-visions contained in Sections 3.8.2. 1. 10 and 6.2.4 of the FSAR wi th allowance =or normal degradation pursuant to the applicable Surveillance Requirements.

5.3 REACTOR CORE FUEL ASSEHBLIES 5.3.1 The reactor core shall contain 217 fuel assemblies with each -;uel assembly con.aining a maximum of 176 fuel rods clad with Zircoloy-4. Each fuel rod sha':1 have a nominal active fuel 1 ngth of 136.7 inches and contain a maxiimum total weight of 2250 grams uranium. Th ini ial core loading shall have a maximum enrichment of 2.83 weight percent U-235. Reload fuel shall be similar in phvsical design to .he initial core loading and shall have maximum enri hment of 3.7 weight percent U-Z35.

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ST. LUCIE - UNIT 1 10/4/79

SAFETY EVALUATION Re: St. Lucie Unit l Docket No. 50-335 Fuel Assembl Enrichment Attachment A: Spent. Fuel Storage Rack Criticality Evaluation Summary Attachment B: New Fuel Storage Criticality Evaluation Summary Attachment C: Fuel Inspection Elevator, Upender, &

Fuel Transfer Tube Criticality Evaluation Summary 10/4/79

ATTACHMENT A Florida Power 8 Light Spent Fuel Pool Storage Rack Criticality Evaluation Summary I. PURPOSE 8 RESULTS This report presents a summary of the criticality evaluation of the high capacity (HI-CAPT<') fuel storage rack designed to acccnodate 728 fuel assemblies in fuel storage locations in the spent fvel pool, at the St. Lucie Nuclear Station, Unit 1 By virtue of the conservative assumptions employed in the criticality evaluation, it is concluded that under. normal operating conditions and with a limiting UOp feed enrichment of 3.7 w/o U-235, the multiplication factor of'lie fully loaded rack in the flooded spent fuel pool does not exceed the limiting multiplication actor of 0.95 specified in A".lSI-i'(210-1976. This conclusion is based on the results of analyses which predict a multiplication factor of 0.947'or the rack when fully loaded with fresh fuel of .the limiting enrichm nt and immersed in pure water at 68oF, including allowances for calculational uncertainties and biases. These analyses eci.ploy arrays of storage cells that are of infinite extent in both the lateral and axial directions, and include the effects of the most adverse combination of mechanical tolerances and fuel displacement. Increasing tne coolant temperature from 68 to 225 F decreases the mul tipl ication factor by 0.015.

DISCUSSIOri This report provides a description of the criticality evalvation of the high capacity fuel storage racks designed and manufactured by Combustion Engineering, Inc. for installation in the speni fuel pool of the St. Lucie lluclear Station, Unit .1. This rack design provides 728 normal storage locations. Each storage location is designed to accommodate one fuel assenbly consisting of 176 fvel rods, and 5 control rod guide tubes in a 14 x 14 array with a nominal pitch of 0.580 inches.

The fuel storage locations are of the HI-CAPTH design >which, for this application, consists of a type 304 stainless steel box structure having a square cross sectional geocetry with a nominal internal di-..ension of 8.4835 inches., The box walls which completely enclose the four vertical sides of a fuel assembly have a nominal thickness o 0.250 inches. The noninat center to center spacing of 12.53 inches for each fuel storag. box, and the nominal water gap thickness of 3.548 inches between. adjacent boxes, are maintained within specified tolerances by structural members welded to the exterior surfaces of the boxes.

Subsequert sections of this report discuss the design bases and results of th cri .icality evalvation.

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A 2 Page 2 III. DESIGN BASES A. Hulti 1 ication Factor The fuel storage rack is designed to meet the subcriti~ality margin specified in Section 5.1.12.1 of ANSI-<<210-1976 which states in part-

"5.1.'2.1- "The spent fuel 'storage racks shall be desioned to assure that a keff not greater than 0,95 is maintained with the racks fully loaded with fuel and flooded with unborated water. The design shall be, based on the maximum enrich..ent.

and fissile isotopic. content oY'uel to be cycled in the plant.

CL B. Assum tions Emolo ed in Criticalit Evaluation The following assumptions are employed in the criticality evaluation to assure that the evaluation is conservative over the range of fuel assembly design variables provided in the specification and/or anticipated operational conditions affecting the criticality margin of the spent fuel pool.

1. Neutron leakage effects are taken to be those characteristic

~ of an infinite array of fully loaded, spent fuel storace locations in the lateral directions and infinitely long fuel assemblies and storage box walls in th axial direction. For the analyses of normal spent fuel locations employing the re erence 8.4835 inch I.O.

'stainless steel box, an infinite array of storage cells having a nominal square dimension of 12.53 inches is employed. In these analyses it is assumed that each fuel storage location contains a fresh fuel assembly of the limiting enr.ichment (3.7 w/o U-235).

2. Parasitic neutron capture contributions in the fuel storage rack structural material are consi rvatively represented by neglecting all structural materials other than the stainless steel box walls.

3. The spent fuel 'pool is assu...ed to be flooded with pure (unborate ') water at a temperature of 68oF. Elevated coolant temperature offects are assessed by'valuating the reac:ivity change between isothermal lattice calcu-lations at 68 and 225 F.

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Page 3 0 A-3

4. Each fuel assembly is assumed to be loaded with unirradiated UOg having an enrichment of 3.7 w/o U-235. No burnable poison pins, control rods, or neutron sources are assumed to be present in the fuel assemblies.

'.,Parasitic neutron capture contributions of structural components in the fuel assembly are conservatively represented by neglecting the zircaloy spacer sleeves and grids.

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6. The, effect of fuel storage rack mechanical tolerances and fuel assenklg;cfisplacement within the fuel. assembly storage box is calculated in a conservative fashion by assuming the ltd'W adverse concurrent combination of dimensional tolerances and a simultanedus diagonal displacement of the fuel assemblies in each cluster of four adjacent storage locations such that each fuel assembly is in contact with two side walls of each box and the spacing between each pair of the four fuel assemblies is minimized. The most adverse concurrent combination of dimensional tolerances corresponds to a configuration wherein the Col'lowing conditions exist in each cell of the storage array: (1) mimimum pitch between centerlines of adjacent fuel storage boxes, (2) maximum storage box internal dimensions, and (3) minimum box wall thickness.

IV. HI-CAP .PACK ANALYSES A general description of the fuel storage rack in the spent fuel pool is given in Section III. The nominal dimensions2of he normal fuel storage locations, defined by CE drawing for the final reference design, are as follows:

I.D. of 304 stainless steel box, in. '.4835 Thickness of'teel box,. in. 0.25 Mater channel, in. 3.548 Center-to-center distance in. 12.5312 The physical parameters for the fuel assembly such as fuel pin radius and density, cell pitch, and ccmposition oi guide tubes are given in Table I.

,The calculated multiplication factor for an infinite array of normal

,fuel storage locations, each containing one fuel assembly centered within th~. stainless steel box, is 0.8984.

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Page 4 A-4 To determine the most adverse effect of mechanical tolerances on the multiplication factor, the extremes in tolerances are used rather than a statistical model. The following tolerances and restraints apply to the nominal dimensions of the final reference design:

I.D. of steel box at top and bottom, in. +0.0625 Hinimum water channel, in. 2 /64 Box wall thickness, in. -0.01. and +0.047 Box wall bow, in. + .250 Center-to-center spacing at top +0.125 bottom from corner of rack, 'nd lne To assess the effect of displacement of fuel assemblies within the storage boxes on the multiplication factor, each fuel assembly is assumed to be displaced diagonally against the corner of its storage box in a direction such that the closest distance of approach is achieved within each cluster of four storage boxes. Rack dimensions are assumed to be those corresponding to the minimum box wall case

~ examined above. The calculated multiplication factor for this case is 0.9324.

To determine the reactivity at 150 F and 225 F for these analyses all materials and dimensions including the csnter-to-csnter spacing were expanded and thermal kernels at 150 F and 225 F were e.:.ployed in,ths cross sections. An additional case at nominal dimensions at 68 F with the more normal 1720 ppm of dissolved boron present was also run.

The last two cases are used to determine the worth of th steel box for an isolated flooded assembly.

The following suttmarizes the results of the seven cases discussed above.

Box C-C Box Mall Case ~Sacin Box E.D. Thickness Yefe Nominal Condition 68 F 12.4375* 8,4835 0.25" 0.8983 2 Hinimum Offset Condition 1'2.0000** 8.6600 0.24" 0.9324 3 Nominal 150 F 12.4473 8.4902 0.25006" 0.8917 4 Nominal 225 F 12.4563 8.4963 0.25011" Oe8836 Nominal 68 F, 1720 ppm boron 12.4375 8.4835 0-25" 0.6751 Isolated, with steel box 32.98" 8.4835 0.25" 0.8075 I

7 Isolated, no steel box 32.98" 0.00" 0.8728

<<The analyzed center-to-center nominal spacing is slightly small er than the constructed value.

<<*Closest fuel assembly center-to-center spacing is 11.46".

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Page 5 The calculational uncertainties used in this evaluation consist of (1) a 0.0053 akeff uncertainty derived from comparisons of calculations for a series of VO~ experirimts, (2) a bias of 0.0019 in overpredicting criticality in the~e experiments, and (3) a bias in the calculating steel box wall worth inferred from calculations of the Johnson-tiewlon experiments3 The magnitude of the latter bias is deduced in the following manner. The worth of the steel box wall structure which is obtained by subtracting th keff of case 6 (0.8075) from that of case 7 (0.8728) is found to be 0.0653 akeff. The analyses of the

. Johnson-Newlon experim nt implied that for the addition of a 0.54 cm thick stainless steel shell to the uranyl fluoride solution container, the worth of steel was overestimat d by a factor of 0.0041 divided by 0.0239 or 0.172., This factor times the calculated worth of steel box walls (0.0653) in the storage rack implies a calculational bias for the steel of 0.0112 ak.

The reactivity balance for the criticality analysis of the normal spent fuel storage locations is summarized as follows:

Host adverse calculated Keff

'.9324

+95/95 confidence level calculational uncertainty 0.0053

+Bias V02 (Experiment - Calculation) -0.0019

'+Stainless Steel Calculational Bias +0.0112 0.9470 Desi n Conditions Hominal l'.ost Adverse Hultiplication Factor for Spen't Fuel Storage Rack 0.8983 0.94?0 Excess Hargin 0.037'1 0.0030 10/4/79

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-References Am rican Nuclear Society, Standards Conmittee horking Group ANS-57.2, "Design Objectives for Light l ater Reactor Spent fuel Storage Facilities at Nuclear Power Stations", ANSI-N210-1976, approved April 12, 1976.

2. CE Drawing .=.E-3077-667-002 Rev. 1, "Spent Fuel Rack llodule".,'ev. l, April 11, 1977.

3., Clark, R. H., et al, Physics Verification Program Final Report BKM-3647-3 (Narch 1967).

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~ II TABLE 1 FUEL ASSEftBLY PARAMETERS Fuel rod pitch. in. 0.58 Fuel rod array Number of fuel rods per assembly 176 Fuel rod clad O.D., in. 0 440 Fuel rod clad I.D., in. 0.384 Fuel rod clad material Zircaloy-4 Fuel pellet diameter, inc .0.3765 Stacked fuel density, gm/cc 10.054 Number of control rod. guide tubes p r assembly Guide tube material Zircaloy-4 Guide tube O.D., in. 1.115 Guide tube I.D., in. 1.035 Fuel Assembly Active F ei Height (in.) 136.7 10/4/79

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~ I ATTACHiviEi)T 8 Florida Power and Light New 5'uel.Storage-Criticality Evaluation Sulggary'URPOSE

& RESULTS'he purpose of this document is to present the results of a criticality evaluation made in lg74 in suppot t of using the St. Lucie-.l ne~ el ~i" storage rack f'r.fresh. Up2 fuels. with.-enrich;:eats. up go 3.7 U235. yfg The new fuel storage racks consist of two arrays of 10 x 4 spaces for fuel assemblies separated by'a 42-inch wide space as shown in Figure 1. The maximum effective neutron multiplication factor under conditions of uniform water (of any density (1 gm/cc) moderation in and between the assemblies should meet the requirements of Section 5.7.4.1 of ANS N18.2 which states:

"The design of spent fuel storage racks and transier equipment shall be such that the effective multiplication facto~

will not exceed 0.95 with new fuel of the highest anticipated enrichment in place assuming flooding ~vith pure'water.

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The design of normaliy'dry new fuel storage racks shall be such that the effective multiplication factor will not exceed 0.98 with fuel of the highest anticipated enrichment in place assuming opiimum moderation (e.g., a uniform density aqueous foam envelopment as the result of'ire figh ing). Credit may be taken for the inherent neutron abs'orbing effect of r.aterials of construction or,, if the requirements of Criterion 5.7.5.10 are met, for added nuclear poisons."

Typically, .his type o, array has a reactivity peak for full density water and a'econdary peak in the v.ater density ringe of 0.03 to 0.2 gm/cc. The rack, although normally dry with a keff of about 0.70, can be immersed in various water densities throuoh ire figh ing foams, floods, etc. For full density water the keff is 0.92, which is vIell within the requirements of a maximum keff of 0.95 including Pnysics uncertainties.

This study uses four-'group transpo."t calculations for water densities ranging betwe n 0.02 and 0.075 gm/cc for the <<w fuel rack, indicatirg a maximum kef of about 0.89.

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B-2 Pag 2 The physics uncertainties are much larger for th se lo:~ density >Iater ystems than for flooded systems, since no applicable exp riments have been performer d. The calculated maximum keff's o about 0. 9 mould allow for an uncertainty of 9.". to meet the 0.98 requirements of Al)S N18.2. This is considered ta be adequate.

Conditions Re uired for Criticalit Safet Th calculations performed indicate that th o.oposed dry storage arrangements meet the criticality safety requ'cerements of NtS Standard HlS 2, with a margin of about 9;l in keff for lo:I density <;ater sideration condi tions. The proposed storage arranger ents are, therefore,"considered to be safe, subject to th folio.~ing conditions:

l. Approved storage racks are used.

2. Criticality safety With plutonium recycle fuel has not been established.

3. The enrichment of the assemblies is limited to 3.70 l

w o U235.

U02 The minimum surface-to-surface spacings bet:~een e

assemblies impl":cit in the analyses, are enforced in the'ack sp cifi-cations (see following section).

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.8-3 Page 3 DISCUSS TON In its evaluation, CE has adopted the approach of assessing s'afety based on the minimum edge-to-edge spacing between any two assemblies, taking into account all dimensional tolerances and anticipated deformations during earthquakes, etc. On this basis, the minimum edge-to-edge spacing between assemblies would be greater than 21.00 - 0.50 (tolerance on pitch) (8 15/16 + 1/16) = 11.5 inches, rather than the 21.00 - 8.2 = 12.8 inches implicit in the analyses.

It is estimated, that the reduction in spacing of 1.3 inches due to tolerances, plus an allowance for defor,.ation (total spacing reduction estimated at 1.5 inches) everywhere would increase tlie

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maximum keff to about 0.916, which would decrease the nargin for Physics uncertainties to about 6... However, the minimum spacing in one position would usually imply a larger spacing elsewhere, and thus the keff could increase to only .901 . The rack design is, there-.ore, judged to provide adequate criticality safety margins for conditions of fog moderation.

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ATTACH~ift<T C Florida Power 8 Light Fuel Inspection Elevator.

Upender, Fuel Transfer Tube Criticality Evaluation Summary Pur ose 8 Results The purpose of this document is to provide a basis for updating Tech Spec 5.3.1 for St. Lucie from an enrichment of .3.1 w/o to 3.7 1 w/o by presenting results of criticality analyses for the fuel inspection elevator, the upender, and the fuel transfer tube.

The applicable standard At<SI-H18.2 (Reference 1) section 5.7.41 states in part "The design of spent-fuel storage racks and transfer equipment shall be such that the effective multi-plication factor will not exceed 0.95 with new fuel of the highest anticipated enrichment in place assuming flooding with pure water,"

The highest reactivity situation', assuming at least a four inch standoff to limit the approach of a second assembly, is 0.911, thus allowing a margin of more than 0.03 ak beyond the allowance for calculational. uncertainties determined by the .analysis of a wide variety of critical experiments.

Desi n Inout The fuel dimensions and densities for the 14 x 14 pin asser,bly are taken from the St. Lucie 1 FSAR using a'3.7 w/o U-235 enrichment.

n, The fuel elevator dimensions are based on Programmed and Remote Systems Corp. Drawing ;.'-15699-0, Rev. 8 dated 6/29/78 of the elevator carriage.

Standoffs appear. in the Ebasco drawing 8770-6841 Rev. 1 to at least partially prevent a second assembly from approaching closer than a ten inch edge to edge separation. The steel structure vias ignored in this analyses.

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. e 0 C-2 The upender dimensions are based on P.R.S.C. Drawinq P A-13594-D Rev. 0 of 1971 for the Fuel Carrier Assembly and indicate that the if,two assemblies are in the carrier assembly, is 4 13/16 inches closest'pproach, and also the presence of four 2x2xl/8 inch stainless steel full length anqles at the corner of each assembly.

The fuel transfer tube inner radius of 35.25 inches was obtained from P.R.S.C. Orawinq "-'l-13~99-0, Rev. E dated 7-28-76 qf the Fuel Trans er Tube Rail assembly Installation.

In all cases non-borated. water at room temperature was assm~ed although normally a few thousand PPl1 of disso1ved boron are present.

Discussion and Results In order to more accurately predict the multiplication factor o, the assembly. arrays, reliable calculations of ti e spatial flux distribu ion,

.esoecially in the neutron absorbing st el regions, are essential. For this reason, a two dimensional transaort calculation mod 1 of the trans-

. fer system is employed in>>hich each component of the fuel transfer system qeometry is explicitly reoresented. Thus, in the fuel upender calculation, the fuel assemblies, the water channel bet:.'een the fuel

- assemblies, the steel angles, and the water reflector are reoresented as separate regions. The fuel assembly itsel= is represented as a 1'.xl4 arrav of fuel assemblv cells containing moderator and either fuel pins or guide tubes. Four neutron group cross sections are genera~ed or e.ch fuel assembly cell and f'r each component of the system with special

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attention qiven to the effect of adjoining regions on the spatial t:". rmal spectrum and hence broad group thermal cross sections of each separate region.

'The most reactive situation of the three considered would be for the fuel elevator when a second fuel assembly is assumed'o be aligned with the one in the elevator with an edge to edqe spacirg of. four inches, i'e resultina keff is 0.911.-

For the uoender the most reactive situation .".ould be when a third asse;..blJ aooroaches to within our inches (edge to edqe) of'he two asse-blies in the upend r; the keff for this situation is 0.899. This kerf is less than for th'e two assemblies seoarated by same distance in the elevator b cause the steel anqles at each corner of both assemblies in the upender were inc1uded in the an'alyses.

The reac.ivity of the fuel array in the trans er tube will be less -..'.an -.or the case o< the upender, i.e. a keff of(0.8'99. The reason being cnlJ two fuel asserblies can be in the transf'er tube and the u~l is maintained in the save confiquratioo as in the upender; a third assembly cannot a"proach the two assemblies while in the transfer tube.

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C-3 The above multiplication factors are valid envelop values for a minimum separation between a third assembly for the upender and a second assembly for the fuel elevator of four inches or greater.

Reference:

l. American National Standards Institute "Nuclear Safety Criteria for the Design of Station and P ressurized Hater Reactor P lants," ANSI-N18.2-1973, August 6, 1973.

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