DESIGN BASIS

HIGH DENSITY FUEL STORAGE SYSTEM AT BROWNS PERRY.

I

I

1. Overall, Descri tion'his'designbasis and safety evaluation report considers the instal-

lation of high density, poisoned, fuel storage racks in the existingBrowns Ferry spent fuel pools. The location of these pools is shown inFSAR Figure 1.6-2 and 1.6-11 (attached).

The e'xisting spent fuel racks have a capacity of 1080 fuel assembliesper pool. 'The high density fuel storage racks will provide a capacityof up to 3471'uel assemblies per pool.

These high density racks are a, base supported modular design .that willreplace, the existing fuel storage and control rod storage racks.Control rod storage will be provided by supplying twenty permanentstorage locations in BF-1 and BF-2 and ei'ghteen in BF-3, and'n aggregateof 370 temporary storage locatioris'. There will be five extra positionsin each pool for storage of defect'ive fuel. The arrangement of thehigh density fuel storage system for„the pools is shown in Figures 1-1,1-2, 1-3, and 1-4.

The high density module provides storage spa'ces for fuel bundles (with-out flow channels) or assemblies (with channels) on approximately 6.5inch center to center spacing. There are two module sizes,. 169 bundle(1'3 x 13) and 221 bundle (13 x 17). The pool capacity of 3471'uelassemblies requires fourteen modules at 13 x 13 and f'ive modules

at'3

x 17.

Each fuel storage module is fabricated from fuel storage tubes, madeby forming"an'outer tube and an inner tube of 304 st'ainless steel withan inner. core of Boral* into a single tube. 'The outer and.inner tubesare welded* together after being sized to the required 'dimensionaltolerances by a patented process. The completed storage tubes are

'astenedtogether by angles welded along the corners and attached to abase plate to form storage modules. Figures 1-6 and 1-7 show schematicallythe 13 x 13 modules. Their overall dimensions are approximately 7 feetsquare and 14, feet high. A 13 x 17 module is approximately 7 feet by9 feet by 14 feet high.

The base plate of each module is support'ed on all four corners by 2-inchthick foot-pads. The foot pads rest on 6-inch thick corner-supportpads which in .turn rest on the fuel pool floor liner. This raises -thebase plate of the module 8-inches above the floor of the fuel pool, allowingsufficient clear area to permit natural circulation of cooling water tothe modules.

April 1978

1

The new spent fuel storage system was designed to conform to thefollowing criteria*.

(1) General Design Criterion 2 as related to componentsimportant to safe'ty being capable of withstanding the.effects of natural phenomena.

V~

~

: .. (2) Ge'neral Design Criterion 3'as relate'd to protect'ion against~

* fire hazard's.

(3) General Design Criterion 4 as related to.components being- able to accommodate the effects of and to be compatible

~, with the environmental conditions associated with normal

o'peration and postulated accidents.

(4) " General Design Criterion 62 as related to the preventionof criticality by physical systems.

(5)

(6)

Regulatory Guide 1.13 as 'it relates to the fuel storage'acilitydesign to prevent damage resulting from the SSE

and to protect the fuel from mechanical damage.f

Regulatory Guide 1.29 as related to the seismic designclassification of facility components.

'7) Regulatory Guide 1.92 as related to combination of loads forseismic analysis.

* General Design Criteria per 10 CFR 50, Appendix A (General Design Criteriafor Nuclear Power Plants) and USNRC Regulatory Guides as noted.

April 1978

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BROWHS =EERY NUCLEAR PLANTFINAL SAFETY ANALYSIS REPORT

Pov/Ic~ Equipment Plans-Es~ons 664 and 639

FIGURE 2.6 2(ReIoed by Amendment 24)

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BROWS FERRY NUCLEAR PLANTFiNAL SAFETY ANALYSIS REPORT

Powerhouse Equipment Plans-Roof and Elevations 664 and 639

FIGURE 1.6-11

0

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Fuel Storage

Defective Fuel Storage

3471

Control Rod StoragePerm. RacksTemp. RackTemp. Hangers

202469

113

STORAGE ARRANGEMENT

BFNP - UNIT 1

FIGURE 1-1

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Fuel Storage

Defective Fuel Storage

3471

Control Rod StoragePerm. RacksTemp. RackTemp. Hangers

202468

TlY

STORAGE ARRANGEMENT

BFNP - UNIT 2

FIGURE 1-2

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Defective Fuel Storage

3471

Control Rod StoragePerm. RackTemp. RacksTemp. Hangers

189491

r03

STORAGE ARRANGEMENT

BFNP - UNIT 3

FIGURE 1-3

,

~,

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0~07

l3< I5 lk< l7 15 < I3 I5)c 13

Note: Plan for Pools No. 2 or 3; Pool No. 1 is similar.

FIGURE 1-4 PLAN DIMENSIONS OF MODULES IN POOL

PUB,L BLBQBPT

LDQ4<cp +TI gc-.~

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I

FIGVRE

FIGURE 1-5 DELETED

April 1978'

7a

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00

t c+JBgcl OSUggPL>$ B,

E)g&Eb~P QBI Q

FIGURE 1~6 ISOMETRIC VIEW OF MODULES

awuppagg pro

p'GAL wgagg+B paaL

C 4c~

tfDoLlL-ENote: Plan for Pools No. 2 and 3; Pool No. 1 js similar.

FIGURE «PLAN OF FUEL STORAGE POOL, SUPPORT PADS AND MODULES

h-9'-

2. Mechanical Structural\

The HDFSS module has been analyzed for both OBE and SSE conditions.Detail stress analysis was then performed to check the design adequacyof'"the module against calculated, loads. R'esults'indicat'ed.that, theHDFSS module design is adequate for ail the combined loacting conditions.

2.1 Seismic Analysis

The HDFSS module has been analyzed for both OBE and SSE conditions.Critical damping ratio of 2X was used in the analysis for both conditions.

. The design floor acceleration response'pectra are given in Figs. 2-3and 2-4. Combination of loads due to-the three components of earthquakeis in accordance with USNRC Regulatory Guide 1.92.

0

The seismic analysis was performed in several steps. First, .the hydro-. dynamic effect, which represents the inertial properties of the fluidsurrounding the submerged modules, was calculated to obtain the hydro-dynamic virtual mass terms based on the module and pool configuration.Three dimensional end effects and leakage between modules are accountedfor by modifying the calculated hydrodynamic mass.

I'igure2-5 shows the plan view of the two-dimensional model of the modulesand pool used in the hydrodynamic virtual mass analysis. The model con-sisted of two rigid bodies: the modules and the pool walls. Waterfinite elements fillthe spaces in between the walls and the modules.The total mass matrix of each module for the analysis is equal to itsstructural mass matrix plus the hydrodynamic mass matrix. Conservativestructural damping values of 2X are applied without any .added dampingdue to fluid effects. The WATER-Ol computer program, GE-proprietary,was used to determine the hydrodynamic mass of one rectangular bodyinside another rectangular body. This program has been design reviewedand meets GE-QA requirements. The

methodologyof calculating hydro

dynamic mass has been presented elsewhere.

(1) L.K. Liu, "Seismic Analysis of the Boiling Water Reactor", Symposium ori~ ~

Seismic Analysis of Pressure Vessel 6 Piping Component, First NationalCongress on Pressure Vessel & Piping, San Francisco, CA, May, 1975.

A ril 1978

-,.10—

PAGE 10 DELETED

April 1978

PROPRIETARY INFORMATION DELETED

It was determined that the lowest fundamental frequency occurred for the13 x 17 module in the 13-tube direction and is equal to 8.7 hz. Thecorresponding frequency for the 13 x 13 module was found to be 9.7 hz.

Table 2-3 lists the results of the sliding analysis, conducted as atwo-dimensional, non-linear analysis, using DRAIN-20 computer code.Note that for the high values of the friction coefficient used, someuplift of the corner was indicated for the model. Since the pulsesobserved by the resultant impact are of short duraction (0.01 to 0.02 sec)and the vertical displacenents are very small; these impact loads willhave only small local effect upon the overall pool slab loading. Pre.—liminary stress analyses indicate that all module stresses in the BrownsFerry modules are less Chan the allowables.

Two cases of the 13 x 17 module half-full were investigated. The casewhere the mass was assumed to be placed on one side of the module causeda small increase in sliding and uplift response, but the results showthan an unsymetrically-loaded module is stable.

0

1

0~ ~

H

Since the SSE maximum floor acceleration exceeds 0;33g, sliding. willoccur and the. maximum base shear force will be limited to the productof,'the submerged weight plus the vertical earthquake effect and -the,coefficient of friction. The OBE peak floor acceleration is equal to0.25g;,however, the lowest .frequency of, the modules was foun'd to be8.7. hz, which corresponds.to a spectral"acceleration of,about 0.5g.Hen'c'e 'it-was found that',there we'll be.a slight amount of sliding;."andthe'maximum OBE base shear force w'ill'be also equal t'o the submerged-weight of the module plus the, vertical earthquake effects times 'thecoefficient of friction.

E

It waa determined that the lowest fundamental frequency occurred forthe 13 x 17 module in the 13-tube direction and is equal td 8.7 hz.The corresponding frequency for the 13 x 13 module was found'to„be,9.7 hz.

Table 2-3 lists the results of the sliding analysis, conducted 'as a

. two-dimensional, non-linear analysis, using DRAIN-2D computei code.This code was originally developed at the University of Californiaat Berkeley, it has been design reviewed and meets NRC QA require-ments. Note that for the high v'alues of the friction coefficient used,some uplift of the corner was indicated for the model. Since the pulsesobserved by the resultant impact are of short duration (0.01 to

0.02'„.sec)

and the vertical displacements are very small, thes'e impact loadswill have only small local effect upon the overall pool slab loading.

r

Two cases'of the 13 x 17 module half-full were investigated. Thecase where the mass was,assumed to be placed on one side of the modulecaused a small increase in sliding and uplift response, but the results

;show that an unsymetrically-loaded module is 'stable.II

2";2 'Str'ess Analysis

The HDFSS module has been designated as seismic category I.; Structural,integrity of the rack has been demonstrated for the load combinations

.be1'ow using linear elastic design methods.'ll module stresses in the'Browns Ferry modules are less than the allowables. Results from stress .

analysis are. presented in Section 2.2.

April 1978

'l

-12--

TABLE 2-1 DELETED

April 1978

- 13-

TABLE 2-2

COMPARISON OF CALCULATED STRESS VS ALLOMABLES (PSI)

OBE CONDITION SSE CONDITION

Location/Type

Tube wall bendiagTube wall shearTube wall tensionTube, weld throat shear

Gale Stress

2,1601,6109,940li560

Allowables

20, 63011,00014,88011,000

Cele Stress

2,4701,84010,5601,790

Allowablesl

41,25022;000'29,76022,000

Angle, weld throat shear 1,390

Casting bending 2,760Casting wall shear 2,850Casting wall compression 9,160

Fuel support plate bending 9,410Support plate weld throat 7,400

bending

Closure plate bending 1,680Closure plate shear 1,750Closure plate weld bending 4,350Closure plate weld shear 1,650

Corner tube local compressive-stress check for local buckling

11,000

20,63011,00016,500

20,63020,630

20,63011,00020,63011,000

1,480

259403,0309,440

9,9907,850

1,9201,8604,9741,750

11,144

22,.000

41,25022,00033~000

41,25041,250

41,25022,00041,250

22,000'7,224

Allowable stresses referenced in ASME Section III, para NF1

April 1978

TABLE 2-3 RESULTS OF NONLINEAR SLIDING AND TILTING ANALYSIS

Ver tical YeduleNodul e/ Earthquake Fuel

~th k C .t' tltl

13 x 17 13-tube direction

Total CoefficientVertical,. of

Force lbs Friction

Naximmn Res nseSliding Uplift

Displacement Displacementin. in.

SSESSESSESSE

SSE

subtractedsubtractedsubtractedadded

SSE none

SSE none

fullfullfullfull

'ne-half full'{symmetric)one-half full(unsymretric)

empty

144,810144,81 0144,810 .

189,240

94.65D

94,650

22,45D

0.100.200.330.33

0.33

0.33

0.10

0.650.280.140.17

0.14

0.37

0.59

0.0000.0030.0190;017

- 0;011

0;;040

0.112

OBEOBEQBE

subtracted fullsubtracted :... fullsubtracted full

155,840155,840155,840

0.100.2Q0.33'.1360.066

0.048

".0.000- 0.0001;0.004

13 x 13

SSESSESSESSESSE

subtractedsubtracted

-subtractednoneadded

fullfullfullfullfull

110,990 .

110,99Q110,990 .-128,000.145,040

0.10. Q.20'0.330.330.33

0.530.230.190.200.18

; 0.'QQQ'.002'::0.01 7

, .0..013—;.0 013

Apiil 1978

~ i

— 15 '-.,

The applied loads, to the rack are:

(a) Dead loads which are'eight of rack and fuel assemblies, andhydrostatic loads.

(b) Live loads, — effect of lifting an empty rack during installation..,C ~

~

II

.(c) .'Thermal loads - the uniform thermal..expansion du'e to pool'emperature,',changes.

I

(d) Seismic forces of OBE and SSE.

(e) Accidental'drop of fuel assembly from maximum possible height.

(f) Postulated stuck fuel assembly causing an upward force of 1000 pounds.

The load combinations considered in the rack design, are:

(1) Live loads

(2) Dead'loads plus OBE

(3)= Dead loads plus SSE

(4), Dead loads plus fuel drop

Stress analyses were made by classical methods for both OBE and SSE .

conditions, based upon the shears and moments developed in the finite-elemen't dynamic analysis of the seismic response. These values werecompared with allowable'tresses referenced in ASME Section III, para.NF (Table 2-2). Values given in Table 2-2 are based on 13 x 17 module.Stresses for the 13 x 13 module are found to. be lower, and therefore;

„. 'not given here. Additional analyses were then performed to determine" the dynamic frequencies, earthquake loading reactions and internal.'-forces in critical module and support system.

The force path in the module due to a horizontal earthquake is shownschematically in Figure 2-1. This figure shows the path of the horizon-tally induced earthquake fuel element inertial forces from the fuelelement to the module support pads. Part of the fuel element inertialforces induced by the motion of the module are transferred eitherthrough the water or directly to the tube walls perpendicular to thedirection of motion (Point 1 in Figure 2-1). These walls then

transfer'he

forces to the side tube walls, which carry the forces down the walls,and into the fuel support plates (Point 2). The portion of the fuelelement load which is not transferred to the fuel tube walls is trans-ferred directly to the fuel support plate at the point where the'owerend fitting of the fuel element is supported vertically (Point 3). Thefuel support plates, acting as a relatively rigid diaphragm, transfer thein-plane shear forces to the long casting which then transfer the shear

, April 1978

15a

3-2

=I QLlE)P I lb.L.L

t=UPL- PLBHBFT

~>sT I2)f=gaL. adl t~gTF'iWTa

L'age<c>apl gc~

HMC)d LBBkaeQH&KQQt

~~f.f INST f ~~

FIGURE 2-1 PATH OF EARTH UAKE HORIZONTAL'FORCES IN NODULE

1

f* = 16.0 cps f = 15.8 cps f = 11.7 cps f = 11.8 cps f = 9.7 cps (13xl3 module)

f = 8.7 cps (13x17 module in13-tube direction)

Fi ni te El ement Fi xed-Base Fi xed-Base Fi xed-BaseInterior Section ll Lumped-Mass ll Lumped-Mass 2 Lumped-Mass

Without Model klithout Model Mith Model With Module withAdded Mass Add'dI Add'dfl Ad dN'« ~5E

*f = Fundamental frequency

FIGURE 2-2 SE UENCE OF MOOULE MOOELS

- 16a—

0oCO

2X Damping

to CO

OoCO

CoCO

'

o ~~

o

CoOl

to CO

loCO

Oo l0 loCO

FAEQVEHCY ~ HERTZ

l toCO

North-South Direct on

Oo00

2X Damping

Ct

cc D.Ot

~ .COEC

Ct to 00

C.CO

I~ CO

Oo 00

Oo lO loCO

FAKOUEHCY ~ HKACZ

East-Mest Direction

FIGURE 2-3 OBE HORIZONTAL EARTH UAKE FLOOR RESPONSE SPECTRA

April 1978

- 16b'-o

0. CO 2X Damping

o

IP~JLJ

Colo

'.IO' r

CC

LJCal loCO

I ~ CO

O.CO

O.IO IoCO IO.CO

FREOUKIICT ~ IIERTZ

North-South Direction

2C Damping

Oo CO

IECA'o00

LJ 0 CO

O.CO

IJA O.CO

IoCO

0. CO

0. I0 loCO

FREOUKIICY ~ 0IKR'IZ

East-Mest Direction

FIGURE 2-4 SSE HORIZONTAL EARTH UAKE FLOOR RESPONSE SPECTRA

April 1978

IllIIIIIII

IIIIll

' ~~ ~ ~

III

IllII

lllIII III

"

III II

III II

III II

p t ~ ~ ~

17

Table 2-4

HIGH DENSITY SPENT FUEL STORAGE SYSTEM

ASSEMBLY DROP ACCIDENT

CASE SUMMARY

No.

2.

3.

4.

Case Descri tion

A fuel assembly drops 18 inchesvertically and impacts the topof a fully loaded YDFSSmodule. The dropped assemblycomes to rest horizontally ontop of the HOFSS.

A fuel assembly drops from 18inches above the HDFSS, entersan empty storage position andfalls to the bottom of thestorage position.

A fuel assembly drops from 18inches above the HDFSS andstrikes a tube wall at anoblique angle.

A fuel assembly drops from 18inches above the top of afully loaded module andstrikes the upper tie plates of2, 3 or 4 fuel assemblies ins.orage.

A fuel assembly drops from 18inches above the HOFSS, falls,outside of the loaded HDFSS,'and'odges adjacent, and parallelto 'an unpoisoned, occupied fuelstorage position.

Effect on Reactivit

Analysis indicates that localized tubedamage or fuel support member damage willoccur, but neutron absorber material willnot be r emoved from its position betweenadjacent fuel assemblies. A fuel assembly,resting horizontally atop the HDFSSdoes not increase the system reactivitybecause the reactivity assumes an infinitevertical length of fuel (no neutron leakagein the vertical dimension). .. keff (0.90

Structural ana lysi s indi ca tes tha t local-ized tube damage will occur. and one neuterabsorber plate may be damaged. A reactiv,~,analysis of this case with the neutronabsorber p'late between two fuel assemblies.totally missing, shows that keff remainsless than 0.90.

It is not possible for a fuel assemblydrop of 18 inches to drive four storedassemblies through the bottom of themodule. Even so, the reactivity effect ofthis impossible event was calculated as alimiting value. An 18 inch section offuel in four bundles in an unpoisoned squararray was found to have a k ff approximatelequal to that of the system. There wouldbe no increase in the overall. reactivity.k ff 0.90.

This case was analyzed for normal handlingconditions. ~ k ff < 0.90.

- 18-

forces to the module base assembly plate (Point 4). The forces arecarried,in the-module base assembly (Point 5) until they are ultimatelytransferred to the module pad and to the support pad and the pool„slab(Point 6).

The path of the vertical forces induced by earthquake motions is somewhatmore complicated. Ultimately, the vertical forces caused by earthquakeand. gravity, loads b'ecome axial forces in .the. module pads. The criticallocation for the-compression forces frcrt'a'the module pads 'is in,the longcastings and tubes'irectly above the module 'pads;-'or stress analysis

'urpos'e,these compressive forces are considered to be resisted by fourfuel tubes sitting directly above the support pad.

Thermal loads were not included in combinations because they were negligible. due to the design of the rack; i.e. the rack is not attached to the structure:.and is free to expand/contract under pool temperature changes. ASME

Section III, Subsection NF, Paragraph'NF-3230 also states that thermalstresses need not be considered. Nevertheless, it was'ound that underthe cooling-water flow conditions specified for the design, the heat risein the storage tube wall due to gamma heat is less than 5 Fat the top ofthe module between tubes and negligible temperature difference at thebottom of the module. Since the module is free to expand laterally at-

'he top of the 'module the thermal stress can be ignored. The maximumwater temperature rise from. the. bottom to the top of a storage tube isabout 16oF. Since the module is free to expand vertically, no thermalstress is generated.

Fuel assembly drop accidents were analyzed. The results are summarizedin Table 2-4.

The loads experienced under a stuck fuel 'assembly condition are less thanthose calculated for the seismic conditions and have therefore not beenincluded .as a load combination.

Analysis was, based upon the criteria'nd assumptions given as follows:

l. ASME Boiler and Pressure Vessel Code Section III, SubsectionNF.

2. USNRC, Reg. Guide 1.92, Combining Modal Responses and SpatialComponents in Seismic'esponse Analysis.

3. Final Safety Analysis Report, Seismic Design Criteria.OBE — Operating Basis EarthquakeSSE — Safe Shutdown Earthquake

4. Seismic Analysis of the High Density Fuel Storage SystemBrowns Ferry Nuclear Power Station — EDAC-134.17 (V5454-1).

5. Light-Gage Cold-Formed Steel Design Manual, 1961 Edition,American Iron & Steel Institute.

April 1978

- 19-

Acceptance criteria were based on:

Normal and upset (OBE) Appendix XVII, ASME, Section III.

,~ z

Faulted .(SSE) Par'a. =F-1370, ASME Section III,„Appendix F..I

'Local bqckling stresses in the spent 'fuel storage tubes were;,- calculated according to "Light-Gage Cold-Formed Steel Design:Manual" of American Iron 6 Steel Institute in lieu of AppendixXVII, ASME, Section III, because of its appliability to theseli'ght-gage tubes.

April 1978

0

03 ~ Materials

- 20-

Most of the structural material used in fabrication of the new HDFSS

is type 304 stainless steel. This material was chosen due to its'orrosion resistance and its ability to be formed and welded with con-sis'tent quality. The 'only material that is not" 304 stainle'ss',steel .;

'."employe'd $ n 'the structure is a special'lloy used between- the foot- '

*- pad and„,support pad,i Soral plates; used as a- neutron absorber, are

an integral non-structural part of. the basic fuel storage tube.. Theseplates are sandwiched between the"inner and outer wall of the storagetube and are not subject to dislocation, deterioration or removal,,deliberate or'nadvertent. The inner and outer walls of the storagetube'are welded together at each end, thereby isolating the Boral" platesfrom direct contact with Spent Fuel Pool,(SFP) water. At normal poolwater operating temperature there is no significant deterioration orcorrosion of stainless steel or Boral.

Specifications were developed specifically for the High Density FuelStorage System which impose requirements to implement and followaccepted and proven industry standards during the design, procurement,fabrication, installation and testing of the storage system. Periodicaudits of the various facilities and practices are performed by certi-fied quality assurance personnel to ensure that these QA/QC r'equirementsare being met. All welding and nondestructive examinat'ion (NDE) isdone in accordance with ASME Boiler '.6 Pressure Vessel Code and theAmerican Society for Nondestructive Testing (ASNT) requirements.

St'orage module components are assembled and welded in special fixturesto maintain a high degree of dimensional tolerance. Each storage positionis checked with full length gauges to assure proper clearance betweenstored fuel bundles and 'storage tube walls.

To provide assurance that specification Boral,sheet is utiliied during,tube fabrication, a'special quality control program is'- in effect at themanufacturer's facility. Samples of each Boral sheet are analyzed todeterm'ine the B10 content. These data are evaluated to verify that th'samples are statistically representative of the entire area of the Boral

~ plate and that B . content, at a 95X" confidence level, meets or exceedsspecification requirements. Analyses are also performed to establish .

the correlation between the B content and the thickness of the Boralsample. The Boral sheets are dimensionally inspected and the thicknessdata are statistically analyzed to verify the sheet meets the'inimumthickness requirement over its entire area at a 95X confidence level.These thickness data are also compared with the correlation data toprovide additional assurance that the B content meets or exceedsspecification requirements. Before each piece of Boral is inserted intoa tube assembly it is verified that each inspection has been successfullyperformed.

Product of Brooks & Perkins, Enc. consisting of a layer of B4C-Al matrixbonded between two layers of aluminum.

April 1978

- 21—

The pr'esence of the neutron absorber material in the fabricated fuelstorage module will be verified at the reactor stoiage-pool site byuse of a neutron source and neutron detectors. There will be apermanent record of all test results that will provide a comparisonbetween the test results for each Boral sheet and the neutron absorption

, rate taken where there. is no Boral sheet.' significant in'crease in.'' the'. neutron'bsorption. rates '.will;verify"the presence 'of 'Boral~;- Module-.- subcriticality calculations have demonstrated Keff <0.95 at, 95X confidence

level with any four complete Boral sheets missing.. A module will beaccepted unless measurements'ndicate that'ive or more Boral sheets arenot present.

Boral has corrosion resistant properties, similar to standard aluminumsheet. Corrosion data and industrial experience confirm that aluminumand Boral have acceptable corrosion resistant properties for theproposed application. Although experience indicates that it is un-necessary, an inservice test program will be conducted, consisting ofperiodically removing and examining, samples of Boral plate which havebeen suspended in the storage pool.

Pool water quality Mill be maintained as'pecified in the BFNP PSAR

section 10.5.4. No changes to water quality are expected as a resultof the planned modification to the spent fuel storage capacity (seesection 3 ' of the environmental assessment).

0USNRC Safety'valuation for Yankee Rove, dated 12/29/76 page 4, Structuraland Material Considerations.

April 1978

, 22

4. Installation and Ins ections *

I. Preparation - Facility

A. Removal of existing storage hardware

l. Disconnect unit connection braces

~ 2. Unbolt hold-down bolts

3. Remove racks from pool

B. Disposal of equipment

l. Decontaminate as required

2. Package for offsite shipment and burial

C. Removal of swing bolts from each foot pad area as required.

1. Remove swing bolts using tools provided

2. Clean all loose material from poolI

1I. Preparation - Equipment

A. Uncrate modules for inspection

l. Use tilting fixture to turn crated. modules fromhorizontal to vertical or upright position

2. Remove crating and visually inspect. Record.

3. Clean as required

4. Attach lifting tool assembly and transfer module torefuel floor

B. Verification of neutron adsorber material existence

l. Disconnect/remove lifting assembly

2. Visually inspect each tube of module for demarcation linesshowing presence of sandwich in each tube wall. Record.

3. Measure wall thickness of 50 randomly selected tube walls.Record.

C. .Xnstall module X.D. strip to perimeter* of each module. I.D.strips must be positioned to preclude damage during installation.

23

III. Installation - Modules

A. Install support pads on pool floor

1. Attach handling fixture to support pad and lower intoposition

2. Align and shim support pad

3. Remove handling fixture

4. Repeat steps 1 through 3 until desired number of padsare in place

B. Installation of modules

1. Attach liftassembly to module

2. 'ower into position and align module (No movement ofmaterials over stored fuel is permitted by the technicalspecifications.)

3. Remove lifting assembly

C. Installation of control rod racks

1. Attach lifting assembly and lower into position

2. Remove lifting assembly

3. Install control rod transfer device

* All handling of heavy loads in the vicinity of the fuel poolswill be accomplished by using the reactor building crane. Theredundant reactor building crane is described in BFNP FSARsection 12.2.2.5.

Nuclear

The following assumptions were used in the analysis of the nuclearCritiCality of the 8/8'telil.

(a) 8 x 8 BWR fuel configuration

(b)

(c)

(d)

{e)

Maximum BWR fuel bundle multipljcation factor (k ) of 1.35in standard core. geometry at 20 C. The use of a%aximum fuelk as a criticality base eliminates the multiplicity ofU-935 enrichment and burnable poison combinations and clarifiesthe exact condition considered.

Storage space pitch of 6.563 in.

Boron { B) equivalent $o a )omogeneous areal concentration of0.01'3 grams (minimum) B/cm .

1

Analysis conservatively performed using 2-dimensional (X,Y)model. (No credit taken for axial neutron leakage).

Credit taken for double wall stainless steel tubes thatseparate fuel bundles.

The criticality analysis calculations were performed with the MERITcomputer program, a Monte Carlo program which solves the neutrontransport equation as an eigenvalue or a fixed source problem includingthe effects of neutron shielding. This program is especially writtenfor the analysis of fuel lattices in thermal nuclear reactors. A geometryof up to three space dimensions and neutron energies between 0 and 10 MeV

can be handled. 'ERIT uses cross sections processed from the ENDF/B-IVlibrary tapes.

The qua) ification of the MERIT program rests upon extensive qualificationstudies including Cross Section Evaluation Work Group (CSEWG) thermalreactor benchmarks (TRX-1, -2, -3, -4) and B&W U02 and Pu0 criticals,Jersey Central experiments, CSEWG fast reactor benchmarks fGOOIVA,JEZEBEL), the KRITZ experiments, and .in addition, comparison withalternate calculational methods. Boron was used as solute in themoderator in the B&W UO~ criticals, and as a solid control curtain in theJersey Central experimeAts. The MERIT qualification program hasestablished a bias of .005 + .002 (1 a ) 6 k with respect to the above criticalexperiments. Therefore, MERIT underpredicts k ff by approximately 0.5percent 6k.

d.C.II .EN i-IVB|Mk1Hi i|F11Spectrum Three Dimensional Monte Carlo Models, ANS Meeting Nov. 1977

25

The storage space (cell) infinite multiplication factor (k ) wascalculated for the high density fuel storage system as deflated bythe assumptions above and the exact geometry specifications. Table 5-1summarizes the results of the k calculations. The maximum k of astorage cell occurs at 20 C witII the fuel bundles centered and IIoflow channels present. Any variation such as increasing the cellpitch, eccentric bundle positionjng, reducing moderator density, andincr easing the temperature to 65 C decreases the k . Table 5-2 showsthe maximum k of the storage cell broken down in% contributingbias and uncerIIainty values. Figure 5-1 demonstrates the decrease in k„with decreasing moderator density. Since the cell is under moderated theoptimum k occurs at 1.0g/cc. The design of the HDFSS has alternatingspaces on Ehe periphery of each module fabricated with unpoisoned closureplates. The unpoisoned locations are also directly opposite each otheron adjacent modules. The effect of the partially unpoisoned storagelocations is small and insensitive to the inter-module water gap asshown in Table 5-3. The maximum module k occurs at the minimum possiblewater gap {1.244") and is less than that of an infinite array of storage cellswith no water gap. The calculational model used to r epresent aninfinite ar ray of modules is shown in Figure 5-2. Each storage cell inFigure 5-2 is a simplified cell (Figure 5-3) used to reduce the requiredHonte Carlo input. The final module k includes the HERIT bias anduncertainty. For the single or simplified cell calculation, no geometrybias would appear due to the fact that MERIT allows an exact representationof the geometry. In conclusion, the HDFSS has a k <.95 at a.95% confidencefor all conditions analysed.

The design of the high density spent fuel storage system has spaces alongthe periphery of the storage pool for storage of defective fuel andcontrol rods. The geometric 'layout is shown in Figure 1 -1, -2, and -3.Analyses have demonstrate'd the HDFSS k c.95 with all peripheral poolstorage locations, including control rod locations, occupied with fuel.Installation accidents that might occur on the refueling floor have been

examined'�

. The worst case accident with no fuel present would be that ofa dropped module. Since the new modules will not be transferred overthe top of either new or spent fuel and will be transferred along apredetermined path, this type accident could result in only damageto noncritical component equipment.

26

TABLE 5-1 - SINGLE CELL HIGH DENSITY FUEL STORAGE CRITICALITY RESULTS

CASE DESCRIPTION

Nominal Rack Dimensions ** With FlowChannel 9 20 C

k ( 2a)"

.8668 + .0075

Nominal Rack Dimensions Without FlowChannel 9 20 C .8674 + .0086

Same as Case 2 except 9 65 C .8561 + .0084

Increased Pi)ch Without flowChannel 9 20 C .8364 + .0106

5 Same as Case 2 but with EccentricBundle Position .8276 + .0123

~includes MERIT Program Bias and Uncertainty at 9BB Confidence Level00

** 6.563" Pitch With Nominal Material Thicknessess

0

27

TABLE 5-2

B IAS & UNCERTAINTY COMPONENTS FOR MAXIMUM Koo OF A STORAGE CELL

00 .8624

'alculation Convergence + .0038

MERIT Bias 8 Uncertainity .005 + .002

Model Bias 5 Uncertainty Hone

Total .8674 + .0086 {2a)

*2acorresponds to 95$ confidence level.

28Figure 5-1

Cell k Versus Moderator Density

X~J

VITIOI

«

«

« ~ q

.90

.80

4 ~

+

.60

.50

4 ~ y r.wo

~ ~

t 1«

.40

' 1

.30

1.0 .8 .6 .4 .2

Moderator Density (g/cc)

29

TABLE 5-3

HDFSS CRITICALITY ANALYSIS

MODULE INTERACTION

Description

Minimum gap between modules

(2A = 1.244 in.)

k (+ 2o)*

.8593 + .0131

Intermediate gap between

modules (2A = 2.100 in) .8579 + .0130

Nominal gap between mo'dules

(2A = 2.967 in.) .8506 + .0134

k includes p~ogram and geometry biases and uncertainties00

at a 955'confidence level.

30

UNPOI SONED

A = 1/2 Gap Between Inside Wallsof Opposing Fuel StorageLocation

FIGURE 5-2

HDFSS MODULE

1/8 MODULE ARRAY FOR NUCLEAR CALCULATIONS

f J

31

H 02

8 x 8 BWR Fuel'er

Assumptions

SS 304

BORAL

SS 304

FIGURE 5-3

HDFSS MODULE

CELL CONFIGURATION MODEL FOR NUCLEAR CALCULATIONS

32

The senstivity of k analyses to various changing parameters areimplied above. Morf specific relationships are as follows:

a ~

b.

Enrichment ercent U-235) - Calculations are based onmaximum, t ere y obviating sensitivity considerations.

00

Stainless steel'hickness - Neutron absorption by the two layersof stain ess stee comprssing the storage tube was includedin the criticality calculations using the nominal thicknesses.The sensitivity of stainless steel within the limits of thethickness tolerances is known but is not significant.

c. Water densit - Figure 5-1 shows the variation of koo withmoderator water ) density.

The effect on reactivity of an accidental fuel assembly drop onto oradjacent to the high density fuel storage system was considered for a

number of postulated cases. The conclusions are presented in Table 2-3.

Our evaluation of the cask tip and cask drop accident is provided in the,Browns Ferry FSAR in the response to NRC question 14.4.

The system k is less than 0.95 at a 95'percent confidence level forany identified seismic or impact loadings.

6. Thermal-H draulic, Total Pool S ent Fuel Coolin

Reactor operations for the Browns Ferry reactors are planned on anannual cycle basis. The high density fuel storage system heat loadfor this annual cycle is based on assumptions that:

o 204 assemblies are discharged per cycle.

o Average exposure is 26,000 NWD/MTU at 23 KW/KgU.

o 8 days cooling (5 days preparation, 3 days unloading) fromreactor shutdown to residence in t'e fuel Rtorspe pool.

A technical feasibility study on the use of an 18-month cycle is beingconducted for Browns Ferry units 1 and 2. If these results are favorable,TVA will evaluate 18-month cycles as a planning basis for all BrownsFerry units. Heat load assumptions for an 18-month cycle are:

o 272 assemblies are discharged per cycle.

o Average exposure is 26,000 NWD/MTU at 23 KW/KgU.

o 8 days cooling as above.

From these assumptions, using the ORIGEN Code*, the heat load per cycle wascalculated. For the annual cycle the7normal heat load is 1.1 x 10 BTU/HR;for the 18-month cycle it is 1.4 x 10 BTU/HR. These values are shownas the ordinate of the first peak on Figures 6-1 and 6-2, respectively.The figures are plots of the batch heat load input and decay between batchesfor the alternative discharge cycles. Discharges at the annual cycle ratewill fi'll the high density system {3471 assemblies), less reserve forone full core (764 assemblies), in thirteen cycles (years). The thirteencycles are shown on Figure 6-1. Similarly, the 18-month cycle willfill the system in ten cycles {15 years) as shown on Figure 6-2.

The maximum heat load was computed by assuming that the total reactorcore is discharged just before a scheduled refueling date. The total coredischarge filled the last available spaces in the pool. Core exposuresat the time of shutdown were in the worst case conditions as given below:

Annual cycle

204 assemblies204 assemblies356 assemblies

1 year exposure - 8,400 NWD/NTU2 year exposure -16,800 MWO/MTU

26,000 NWO/MTU

Bell, M.J., "ORIGEN Code - The ORNL Isotope Generation and Depletion",ORNL-4628

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-6-

36

18-month cycle

272 assemblies492 assemblies

1.5 year exposure - .12,600 NWD/MTV26,000 MWD/MTU

It was further assumed that sixteen days cooling time was required tounload all of the fuel (5 days preparation and ll days unloading).

Heat load results are shown in Figure 6-3. From reactor shutdown tothe end of the discharge period, the heat. load decreased to approximately28 million BTV/HR.

Water flow through the passages in the storage module and supportsystem w'ill be adequate to maintain cooling with the heat transfersystems available. With the incoming water to the fuel s~tora e poolat 125 F, the maximum fuel cladding temperature will be 165op,The maximum water temperature associated with the hottest fuel bundlewiU be 144'P. Those temperatures are low relative to structuralintegrity or corrosion limiting temperatur es of the structur al componentsof the storage system and fuel.

Continuing efficiency of the exchange of heat from the spent fuel to thepool water depends on the convection flow of water through the storagelocation and flow channel encompassing a fuel bundle. The floe-like crudthat adheres to the surfaces of the spent fuel bundle was studied todetermine whether it is a potential mechanism for blocking flow through theChannel. It WaS fOund that the floe iS eXtremely fine, SuCh that pieCeSthat spall off of the aggregate are not disposed to settle, but may flowupward with Che convection current. Additionally, the floe"is so fine thatsome of it will pass through conventional laboratory filter papers.Growth of crud in fuel storage conditions has not been observed in commercialfacilities. The potential for channel plugging by sedimentation or byblockage of flow passages is negligible.

The high density fuel storage system and the BWR fuel to be stored in it arenot fabricated such Chat significant quantities of air or other gas can beentrapped creating an area of reduced effective moderator density ..Buteveo if. air were trapped, the effect of reduced density on the under-moderated fuel bundles is to reduce the k ff of the system.

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38

As previuosly described that the maximum water temperature in the system(based on incoming water temperature of 125'F) is 144'F. From thisit is apparent that there is no possibility of boiling in the pool,thereby eliminating steam formation as a source of variations in moderatordensity.

Two loss of cooling cases have been analyzed to indicate the relativetemperature inertia of the expanded pool compared to its existing capacity.

The first case is that of the full core discharge with the pool full offuel discharged under the normal schedule except for the slots reservedfor the full core unload. The worst condition, represented by the fuelexposures of the 18-month cycle, is used in this case. Cooling is lost atthe completion of the discharge (sixteen days) with the pool temperatureat 150'F.

For the existing capacity, Case 2 assumes that one discharge of 272 bundles isalready in the pool. At the time of the next discharge, the total core isdischarged making a total of 1036 bundles (existing capacity is 1080). Theexposure of the full core and initial pool temperatures are the same as thosein (use 1.

The calculational method used allows for evaporation from the pool surface,but not conduction through the pool walls. Results of the calculations areplotted in Figure 6-4.

The existing fuel pool cooling system is described in detail in FSAR section 10.5.As described in the FSAR (section 10.5.5) the RHR system is operated inparallel with the fuel pool cooling system and acts as a seismic Category Ibackup. No modifications are planned to be made to the existing pool coolingsystem because calculations have proven that the existing system has thecapability of handling the expected heat load.

The fuel pool cooling system is a redundant system consisting of two heatexchangers and two pumps with separate power supplies. In the unlikelyevent that both of these are inoperative or the heat load exceeds the capacityof the, spent fuel cooling system, the RHR system may be used as a backup.Indication to the operator of any loss of cooling capacity is provided bythe spent fuel pool cooling system and RHR, system parameters monitored inthe main control room. Additionally, the temperature of the spent fuel poolwater is recorded on a scheduled basis as required by technical specification.Should cooling be lost for an extended period of time such that the temperatureof the pool became elevated sufficiently to cause evaporation of the poolwater, makeup to the pool could be supplied by means of a fire hose. Thefuel pool itself is so designed to preclude inadvertent draining of the pool.

NO, 3}O SO 4}VISIONS RXR IHCH OOYH wAYS ~ SO ST SOO OIVIt'OMS. t@Q~~O}H STOCK OIRSCT SROH COOEX ROOK CO HORWOOO, MASS CROSS

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ENVIRONMENTAL ASSESSMENT

1.1 STORAGE

Although TVA's fuel supplier for Browns Ferry is required under

the terms of its contract to remove and reprocess discharged spent

fuel, the current absence of reprocessing industry has necessitated

the storage of discharged spent fuel. TVA has agreed with the fuel

supplier to increase the Browns Ferry storage capacity in order to

accommodate the spent fuel until it can be reprocessed or otherwise

disposed of.

The original capacity of each spent fuel pool was 1080 fuel

assemblies. The unit 1 and 2 pools are connected by a spent fuel

transfer canal: There is no such interconnection fox the unit 3

pool which therefore has a more limited storage capacity. On the

basis of maintaining a full core reserve storage capability in the

pool serving Browns Ferry 3 and one-half full core reserve storage

in each of the 'pools serving Browns Ferry 1 and 2, the storage

capability available'ox use is 316 fuel assemblies in the pool

serving Browns Ferry 3 and 698 fuel assemblies in each of the pools

serving Browns Ferry 1 and 2. Storage capacity with full core dis-

charge capability would therefore be exceeded in 1980 for Browns Perry 1,

in 1981 for Browns Ferry 2, and in 1979 for Browns Ferry 3.

The proposed expansion provides storage for all discharges through

1991 for Browns Perry 1, through 1992 for Browns Ferry 2, and thxough

1990 for Browns Ferry 3, while maintaining the full core xeserve

storage capability as described previously. Therefore, storage capacity

is extended for ll years for each of the units. In addition, five

defective fuel assembly storage positions are provided for the storage

of leaking or grossly defective fuel assemblies in the event they are

required.

1.1.1 REFUELING SCHEDULES

The total storage capacity expected to be utilized is based on

maintaining a full core reserve storage capability available to

each Browns Ferry reactor. Since the pools serving Browns Ferry 1

and 2 are connected so as to permit spent fuel transfers from one

pool to another without a cask, a shared fulI. core reserve storage

capability is expected to be maintained with each pool contributing

approximately one-half of the reserve capability. The estimated

refueling schedules and expected number of fuel assemblies to be

transferred into the spent fuel pools are given in the following

tables.

Browns Ferr 1

RefuelingDate

Number of FuelAssemblies Dischar ed

Cumulative Numberof Fuel Assemblies

in SFP

Sept. 1977Sept. 1978Sept. 1979Sept. 1980Sept. 1981Sept. 1982Sept. 1983Sept. 1984Sept. 1985Sept. 1986Sept. 1987Sept. 1988Sept. 1989Sept. 1990Sept. 1991Sept. 1992Sept. 1993

168220196196204200200200200200200200200200200200200

168388584780984

1184138415841784198421842384258427842984

{Storage Limit)... 3089(Storage Limit)... 3089

Brogans Fcr

refuelingDate

March 1970March 1979March 1980March 1901March 1982March 1983March 1984March 1985March 1986March 1987March 1988Narc+ 1989March 1990March 1991March 1992March 1993March 1994

Number of Fuel'Assemblies Dischar ec1

1602201961962o4200200200200200200200200200200200200

Cumulative Number of FuelAssemblies in SFP

16038050478o984

11841384158417841984218423842584278429843089 (Storage Limit)3009 (Storage Limit)

Browns Fer 3

RefuelingDate

Sept. 1978Sept. 1979Sept. 1980Sept. 1981Sept. 1982Sept. 1903Sept. 1984Sept. 1905Sept. 1986Sept. 1907Sept. 1988Sept. 1989Sept. 1990Sept. 1991Sept. 1992

Number of FuelAssemblies Dischar ed

2o02o0108,200200200200200200200200200200200200

Cumulative Number of FuelAssemblies in SPP

2o84166o4804

loo412o41404160410o4200422o4240426042707 (Storage Limit)270'7 (Storage Limit)

After the first refueling of a Browns Ferry reactor Curing the fall of 1977,

168 fuel assemblies vi11 be stored. in the spent fuel pool serving the

Brows Ferry 1 reactor.

II

l. 2 HEAT ADDITION

The incremental heat load increase resulting from the proposed

modification is shown graphically on Figures 6-1 and 6-2 of the

Design Basis Report. The modified basin capacity is 3471 bundles,

equivalent to 17 annual discharges or 13 discharges at 18months'ntervals.

The existing capacity is 1080 bundles; 5 annual discharges

or 4 discharges at 18 months. The heat load increment of the expanded

7fuel storage over existing storage is 0.15 x 10 Btu/hr. The heat

loads are tabulated below:

TOTAL POOL PEAK HEAT LOAD-7

HEAT LOAD BTU/HR x 10

~Ex ended ~Exkstdn Increment

Annual Cycle

18-Month Cycle

1.48

1.78

1.33

1.63

0.15

0.15

With the heat removal capacity available for cooling the spent fuel

pool water, increased water temperatures and evaporation rates will be

small. The plant cooling water system will accommodate the additional

heat load. The increase of heat load contribution of stored spent fuel

to total plant thermal discharge to the environment during normal operation

is less than 0.02 percent.

2.0 COST

The total cost associated with the project for all three Browns Ferry

units is expected to be about $ 19 million in 1977 dollars. This

estimate includes the, following five categories of expense:

1. Project management, design, quality assurance, and licensing.

2. Materials, tooling, and hardware fabrication.

3. Removal, installation, and transportation.

4. Contingency allowance.

5. Allowance for funds used during construction.

3.0 RADIOLOGICAL EVALUATION

3.1 SOLID WASTE

A Fuel Pool Cooling and Cleanup System (FPCCS) is provided for each

unit and is described in BFNP FSAR section 10.5. Removal of impurities

of the fuel pool water is accomplished by three pressure-precoat type

filter-demineralizers with a fourth as a spare. Each filter-demineral-

izer has a flow capacity greater than or equal to the system design flow-

rate. The maximum system Xlowrate is twice the flowrate required to

maintain pool water quality. The amount of dewatered resin shipped from

BFNP is about 2300 cubic feet per month. The volume of dewatered resin

from the FPCCS is 0.5 percent of the plant total dewatered resin.

Operating experience has shown that the changeout rate of the precoat

resins is determined by water clarity requirements and not by radio-

nuclide concentrations. The filter-demineralizers will maintain the

concentration of radioactivity in the spent fuel pools below 0.01 p Ci/cc

of Cs-137 equivalent regardless of the number of fuel assemblies stored

in the pools. As the number of stored assemblies increases, the interval

between required precoat resin changes may decrease slightly, however,

changes are expected to continue to be made on a monthly basis. The

volume of solid wastes shipped offsite due to the increased spent fuel

storage capacity is not expected to increase.

4

3.2 RADXOXSOTOPE INCREASE

Radioactive materials can be released to the fuel pool water from

leaking fuel assemblies. Non-volatile material mill remain in the

pools while noble gases will be released to the building atmosphere.

The long term storage of additional fuel will provide significant time

for decay of fission and daughter products. Even between consecutive

refueling outages, any increases in radioisotopes due to additional

stored fuel will consist only of long-lived isotopes.

The only long-lived radioactive noble gas of significance is K-85.

After a refueling discharge batch has cooled in a pool for 12-18 months,

the driving mechanism for release of additional amounts of K-85 has!

become very small. A conservatively estimated additional amount of

K-85 that could potentially be released will be small compaxed to the

total annual quantity of all noble gases released from the pools and

negligible when compared to the total annual plant noble gas releases.

3.3 SHXELDXNG

The spent fuel is shielded by more than 21 feet of water. Because

of this depth of water, radiation levels at the pool surface are

controlled by the radionuclide concentrations in the pool water. The

concentrations of radionuclides in the pool water will be maintained

below design levels regardless of the number of fuel assemblies in

the pool. Therefore, radiation levels vill not increase at the

pool surface with increasing quantitios of stored pent fuel.

Most of the walls of the spent fuel pools are five feet thick

concrete and narrow to four and one-half feet in a few places.

The floors of the spent fuel pools are five feet thick. Design

radiation levels outside these walls and the floor will not be

exceeded by increasing the quantity of stored spent fuel or by

installing the new racks as close as 18 inches 'from the walls

inside the pools.

3.4 0 ERATIONAL EXPOSURE CHANGES DURING NODAL OPERATION

The man-rem accumulation during normal plant operation wil1 not be

affected by the additional fuel assemblies. Three areas of plant

operation considered in this evaluation are filter-demineralizer

resin changes, personnel occupancy of pool'reas, and refueling floor

airborne activity levels.

The changeout of the filter«demineralizer precoat resins willcontinue

to occur at monthly intervals. This is because the changeout rate will

be controlled by water clarity requirements and not by radionuclide

concentrations. The backflushing operation is controlled remotely from

the Radwaste Building and plant personnel do not physically approach the

filter-demineralizers, valves, or pumps.

Since the filter-demineralizers will maintain the concentrations of radio-

nuclide below 0.01 p Ci/cc of Cs-137 equivalent regardless of the number

of spent fuel assemblies stored in the pool, radiation levels at the pool

surface will not increase. In addition, routine radiation'surveys over

the periods of operation and storage of spent fuel at Browns Ferry

do not indicate any trend for radioactive crud buildup on the sides

of the pool. Information from other utilities with operating plants

also indicates that crud buildup should not be a problem.

The levels of airborne radioactive materials around the pool other than

noble gases will not increase because of the cleanup capacity of the

FPCCS. The small potential increase in the levels of the long-lived

radioactive noble gas K-85 will have a negligible affect on doses to

personnel in the Reactor Building. Therefore, routine refueling

activities and normal personnel occupancy of the pool areas will not add

to the plant man-rem burden with additional stored spent fuel.

3.'5 „OCCUPATIONAL EXPOSURES DUE TO CHANGEOUT OF SPENT FUEL RACKS

The spent fuel pool in unit 3 has never contained spent fuel. The old

racks in unit 3 will not contribute to the occupational exposures

accumulated during the changeout. The changeout of the racks for unit 3

is planned to occur while unit 3 is at power. Low level dose rates may

occur, near the refueling slot shield plugs between the reactor refueling

cavity and the spent fuel pit. Precautions will be taken to limit the

amount of time spent in this area in order to minimize personnel exposures.

The initial work of installing 7 modules on unit 3 is planned to be done

in a dry pool before any fuel is discharged to the pool. (However,„ ifthis work is done with fuel in the pool, the water present will provide

sufficient shielding and associated operations will be performed with

long-handled tools.) The final installation of the remsining rack,mo'dules

will be accomplished underwater after the first refueling. Occupational

-9-

exposures for this work will be limited primarily to decontamination

of tools and exposure to the fuel pool water. There are 54 aluminum

original racks in unit 3 with a total weight of 103,950 pounds. The

uncontaminated racks will be disposed of as scrap.

The unit 1 and 2 pools will contain spent fuel. The unit 2 racks will

be changed after the initial work is finished on unit 3. The unit 2

spent fuel will be stored temporarily in the unit 1 pool while work

is going on in the unit 2 pool. The spent fuel will be moved through

the fuel transfer slot which connects the unit 1 and 2 pools. It is

anticipated that 168 assemblies will be moved. Measurements taken

at unit 1 have shown that dose rates at the fuel handling bridge while

transferring fuel assemblies do not exceed 2.0 mr/hr. Assuming an

average of 10 minutes for each fuel assembly, the total man-rem

accumulated during this operation should not exceed 0.1.

The pool will be drained and decontaminated. The racks will be

decontaminated, crated, and shipped offsite to a licensed burial

location. Every reasonable effort will be made to limit personnel

exposures to as low as is reasonably achievable during this work.

There are 54 aluminum racks in unit 2 with a total weight of

103,950 pounds.

The final installation of the remaining rack modules in unit 2 will

be accomplished after the unit 2 spent fuel has been moved back into

the unit 2 pool and will be done underwater. Occupational exposures

for this work will be similar to those for the final underwater

installation of unit 3.

The operation for the changeout of the unit 1 racks will be similar

to the unit 2 changes. There will be about 600 fuel assemblies in

the unit 1 pool which must be transferred to the unit 2 pool through

the connecting fuel transfer slot resulting in personnel exposures of

less than 0.3 man-rem. Removal and disposal procedures followed willbe similar to those followed on unit 2. Experience gained through

the work on units 3 and 2 should result in a reduction of personnel

exposures for the unit 1 work as compared to the unit 2 work. There

are 54 aluminum original racks in unit 1 with a total weight of

103,950 pounds,

For the complete change of racks for the three units and adding

ten percent for miscellaneous material, there will be 114,345 pounds

of aluminum disposed of as scrap and 228,690 pounds shipped to an

offsite, licensed burial area.

An informal survey of operating BWR plants indicates that the ma)or

contributor to personnel exposures during rack changeouts is the

re@oval and decontamination of the old racks. Indications are that

the changeout work will result in total man-rem accumulations of

about 20-30 per unit for units 1 and 2 and less than five man-rem for

unit 3. It is anticipated that this will amount to less than 10

percent of";the annual man-rem for each unit after one refueling has

been accomplished.

J

3.5.1 OPPSITE DOSES

There will be no significant increase in either liquid or gaseous

effluents as a result of increased spent fuel storage; therefore,,

there should be no detectable increase in offsite doses.

4. 0 RESOURCE COMMITMENT

The relatively small quantities of material resources being committed

would not significantly foreclose the alternatives with respect to

other licensing actions designed to ameliorate a possible shortage of

spent fuel storage capacity. The principal material resources that willbe consumed by the proposed modification together with estimated annual

domestic consumption are indicated below.k

Material

304 Stainless Steel

Boron Carbide

Aluminum

Browns Ferry Modificationuantit lbs.

1.12 x 10 6

2.71 x 10

1.25 x 105

Annual U.S.Consum tion lbs.

2.82 x 10

3 to 9 x 10

8 x 10

5.0 ALTERNATIVES

W

Although the TVA fuel supplier has the responsibility for disposition

of the spent fuel, the alternatives to the proposed modification which

have been considered by TVA are:

1. Shutting down the Browns Ferry reactors for lack of spent

fuel storage capability.

2. Shipping spent fuel to a facility for reprocessing.

-12-

3. Shipping the spent fuel to an independent offsite storage

facility.4. Shipping the spent fuel to another reactor-site spent fuel pool.

5. Shipping the spent fuel to a waste repository.

The first alternative is unacceptable relative to the proposed

modification. Replacement power {ifavailable at all) is expected to

cost an average of at least 16 mills per kilowatt-hour greater than

the cost of generation from the Browns Ferry reactors. Shutting down

one reactor is estimated to result in additional costs of at least

$ 9 million per month. Replacement of the generating capability that

would be lost by shutting down the Browns Ferry reactors would be

many times more expensive than the proposed modification.

The second alternative is not now available and is not expected to

become available in the near future in view of 'the President's

proposal to postpone reprocessing indefinitely and in view of the

long lead times (on the order of 10 years) required to plan and

construct facilities for the reprocessing of spent fuel. The need

for storage capacity would exist even if governmental policy immediately

allowed reprocessing of spent fuel.

The third alternative is not now feasible. The offsite storage

facilities now in existence are inadequate to meet the near-term demands

of the industry, and it is very unlikely that offsite storage facilities

could be constructed on a schedule that would eliminate the need to

expand storage capacity in the reactor site spent fuel pools. The cost

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of storage at an offsite facility would be considerably more

expensive than providing storage at the reactor site spent fuel

pools. An independent facility would possibly require acquisition

of additional land and would necessarily require construction of a

spent fuel pool with associated containment, purchase of heat removal

systems, shipping cask and spent fuel transportation system, plus

operational and security personnel. The proposed modification requires

only the installation of spent fuel storage racks. Obviously the third

alternative is much more expensive than the proposed modification.

The fourth alternative requires the transport of spent fuel to another

reactor. A reactor with a pool available for storing fuel discharged

from Browns Ferry would either be operating or very near to operation.

The need for additional storage capability would only be delayed a

relatively short period of time by transporting to another reactor site

spent fuel pool. The receiving reactor pool would become more quickly

filled with spent fuel thus very likely making it necessary to find

storage gust a few years later by much more expensive means. The

transportation cost to another reactor site would also be substantial'I

compared to the costs of the proposed modifications for the Browns Ferry

spent fuel pool. Additionally, no reactor site to accept fuel from

Browns Ferry has been identified. Other TVA reactor site spent fuel

p ools are not designed to accept fuel of the Browns Ferry design or

would not be completed in time to receive fuel from Browns Ferry.

Other reactor owners are facing situations similar to TVA with respect

to storing spent fuel and therefore do not desire to store fuel from

TVA reactors. This alternative is therefore undesirable and also is

not known to exist.

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The fifth alternative does not now exist and there are no plans

for a waste repository to be constructed on a schedule that would

eliminate the need for interim storage to accomodate the near term

discharges from Browns Perry. The Federal Government has plans to

build pilot facilities by 1985 for the demonstration of disposal

concepts. Facilities of a larger scale would very likely follow

several years later. This alternative is therefore unavailable on

a suitable schedule.,

In conclusion, the proposed modification is considered the most

desirable and economical of the alternatives considered.

6.0 REFUELING ACCIDENTS

For the radiological effects of the cask drop test, reference the

environmental statement volume No. 3, section 2.3.3(b), pages 2-42,

2-46, and 2-47, and Table 2.3-2.

Since only fuel bundles are to be transferred over the spent fuel,

we do not see the need for an addition to the technical specifications;

i.e., the maximum load transported over the spent fuel will be that of

a single fuel assembly.

7.0 CONCLUSIONS

The alternatives described above do not offer the operating flexibility

of the proposed action nor could most of them be completed as rapidly

as the proposed action. The alternatives of shipping the spent fuel to

t

a reprocessing facility, an independent storage facility, or to

another reactor would be more „expensive than the proposed action

and either of these alternatives might pre-emp storage space needed

by another utility. The alternative of ceasing operation of the

facility also would be more expensive than the proposed action because

of the need to provide replacement power. In addition to the economic

advantages of the proposed action, we have determined that the expansion

of the SFP would have a negligible environmental impact. Accordingly,r

deferral or severe restriction of the action here proposed would result

in subst'antial harm to the public interest.

Based on the results of this environmental assessment, i.e., negligible

increase in offsite doses and negligible increase in doses to personnel

from radionuclide concentrations in the SFP, the conclusions and

determinations of the Browns Ferry Final Environmental Statement have

not significantly changed.

+ A

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