ML18283B548

Design Basis High Density Fuel Storage System at Browns Ferry

ML18283B548
Person / Time
Site:	Browns Ferry
Issue date:	04/30/1978
From:	Tennessee Valley Authority
To:	Office of Nuclear Reactor Regulation
References
Download: ML18283B548 (99)

Text

DESIGN BASIS HIGH DENSITY FUEL STORAGE SYSTEM AT BROWNS PERRY.

I I

1.

Overall, Descri tion

'his'design basis and safety evaluation report considers the instal-lation of high density,

poisoned, fuel storage racks in the existing Browns Ferry spent fuel pools.

The location of these pools is shown in FSAR Figure 1.6-2 and 1.6-11 (attached).

The e'xisting spent fuel racks have a capacity of 1080 fuel assemblies per pool.

'The high density fuel storage racks will provide a capacity of up to 3471'uel assemblies per pool.

These high density racks are a, base supported modular design.that will replace, the existing fuel storage and control rod storage racks.

Control rod storage will be provided by supplying twenty permanent storage locations in BF-1 and BF-2 and ei'ghteen in BF-3, and'n aggregate of 370 temporary storage locatioris'.

There will be five extra positions in each pool for storage of defect'ive fuel.

The arrangement of the high density fuel storage system forthe 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 6.5 inch 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'uel assemblies 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, made by forming"an'outer tube and an inner tube of 304 st'ainless steel with an inner. core of Boral* into a single tube.

'The outer and.inner tubes are welded* together after being sized to the required 'dimensional tolerances by a patented process.

The completed storage tubes are

'astened together by angles welded along the corners and attached to a base plate to form storage modules.

Figures 1-6 and 1-7 show schematically the 13 x 13 modules.

Their overall dimensions are approximately 7 feet square and 14, feet high.

A 13 x 17 module is approximately 7 feet by 9 feet by 14 feet high.

The base plate of each module is support'ed on all four corners by 2-inch thick foot-pads.

The foot pads rest on 6-inch thick corner-support pads which in.turn rest on the fuel pool floor liner.

This raises

-the base plate of the module 8-inches above the floor of the fuel pool, allowing sufficient clear area to permit natural circulation of cooling water to the modules.

April 1978

1

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

(1)

General Design Criterion 2 as related to components important 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 prevention of criticality by physical systems.

(5)

(6)

Regulatory Guide 1.13 as 'it relates to the fuel storage

'acility design 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 design classification of facility components.

'7)

Regulatory Guide 1.92 as related to combination of loads for seismic analysis.

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

April 1978

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PLAN-fL699Lf BROWHS =EERY NUCLEAR PLANT FINAL SAFETY ANALYSIS REPORT Pov/Ic~ Equipment Plans-Es~ons 664 and 639 FIGURE 2.6 2 (ReIoed by Amendment 24)

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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 adequacy of'"the module against calculated, loads.

R'esults'indicat'ed.that, the HDFSS 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. 2-3 and 2-4.

Combination of loads due to-the three components of earthquake is 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 fluid surrounding 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 accounted for by modifying the calculated hydrodynamic mass.

I'igure 2-5 shows the plan view of the two-dimensional model of the modules and pool used in the hydrodynamic virtual mass analysis.

The model con-sisted of two rigid bodies:

the modules and the pool walls.

Water finite elements fillthe spaces in between the walls and the modules.

The total mass matrix of each module for the analysis is equal to its structural mass matrix plus the hydrodynamic mass matrix.

Conservative structural damping values of 2X are applied without any.added damping due to fluid effects.

The WATER-Ol computer program, GE-proprietary, was used to determine the hydrodynamic mass of one rectangular body inside another rectangular body.

This program has been design reviewed and meets GE-QA requirements.

The methodology of 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 National Congress 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 the 13 x 17 module in the 13-tube direction and is equal to 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-20 computer code.

Note that for the high values of the friction coefficient used, some uplift of the corner was indicated for the model.

Since the pulses observed by the resultant impact are of short duraction (0.01 to 0.02 sec) and the vertical displacenents are very small; these impact loads will have only small local effect upon the overall pool slab loading.

Pre.

liminary stress analyses indicate that all module stresses in the Browns Ferry modules are less Chan the allowables.

Two cases of the 13 x 17 module half-full were investigated.

The case where the mass was assumed to be placed on one side of the module caused a small increase in sliding and uplift response, but the results show than an unsymetrically-loaded module is stable.

0

1 0

~

~

H Since the SSE maximum floor acceleration exceeds 0;33g, sliding.will occur and the. maximum base shear force will be limited to the product of,'the submerged weight plus the vertical earthquake effect and -the, coefficient of friction.

The OBE peak floor acceleration is equal to 0.25g;,however, the lowest.frequency of,the modules was foun'd to be 8.7. 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;."and the'maximum OBE base shear force w'ill'be also equal t'o the submerged-weight of the module plus the,vertical earthquake effects times 'the coefficient of friction.

E It waa determined that the lowest fundamental frequency occurred for the 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'tobe, 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 California at 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 pulses observed by the resultant impact are of short duration (0.01 to 0.02'.sec) and the vertical displacements are very small, thes'e impact loads will have only small local effect upon the overall pool slab loading.

r Two cases'of the 13 x 17 module half-full were investigated.

The case where the mass was,assumed to be placed on one side of the module caused 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 TABLE 2-1 DELETED April 1978

TABLE 2-2 COMPARISON OF CALCULATED STRESS VS ALLOMABLES (PSI)

OBE CONDITION SSE CONDITION Location/Type Tube wall bendiag Tube wall shear Tube wall tension Tube, weld throat shear Gale Stress 2,160 1,610 9,940 li560 Allowables 20, 630 11,000 14,880 11,000 Cele Stress 2,470 1,840 10,560 1,790 Allowablesl 41,250 22;000

'29,760 22,000 Angle, weld throat shear 1,390 Casting bending 2,760 Casting wall shear 2,850 Casting wall compression 9,160 Fuel support plate bending 9,410 Support plate weld throat 7,400 bending Closure plate bending 1,680 Closure plate shear 1,750 Closure plate weld bending 4,350 Closure plate weld shear 1,650 Corner tube local compressive-stress check for local buckling 11,000 20,630 11,000 16,500 20,630 20,630 20,630 11,000 20,630 11,000 1,480 25940 3,030 9,440 9,990 7,850 1,920 1,860 4,974 1,750 11,144 22,.000 41,250 22,000 33~000 41,250 41,250 41,250 22,000 41,250 22,000'7,224 Allowable stresses referenced in ASME Section III, para NF 1

April 1978

TABLE 2-3 RESULTS OF NONLINEAR SLIDING AND TILTING ANALYSIS Ver tical Yedule Nodul e/

Earthquake Fuel

~th k

C t'

tltl 13 x 17 13-tube direction Total Coefficient Vertical,.

of Force lbs Friction Naximmn Res nse Sliding Uplift Displacement Displacement in.

in.

SSE SSE SSE SSE SSE subtracted subtracted subtracted added SSE none SSE none full full full full

'ne-half full

'{symmetric) one-half full (unsymretric) empty 144,810 144,81 0 144,810 189,240 94.65D 94,650 22,45D 0.10 0.20 0.33 0.33 0.33 0.33 0.10 0.65 0.28 0.14 0.17 0.14 0.37 0.59 0.000 0.003 0.019 0;017

- 0;011 0;;040 0.112 OBE OBE QBE subtracted full subtracted

... full subtracted full 155,840 155,840 155,840 0.10 0.2Q0.33'.136 0.066 0.048

".0.000

- 0.0001

0.004 13 x 13 SSE SSE SSE SSE SSE subtracted subtracted

-subtracted none added full full full full full 110,990 110,99Q 110,990.-

128,000.

145,040 0.10

. Q.20

'0.33 0.33 0.33 0.53 0.23 0.19 0.20 0.18

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, and hydrostatic 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 were compared 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 shown schematically in Figure 2-1.

This figure shows the path of the horizon-tally induced earthquake fuel element inertial forces from the fuel element to the module support pads.

Part of the fuel element inertial forces induced by the motion of the module are transferred either through the water or directly to the tube walls perpendicular to the direction 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 fuel element load which is not transferred to the fuel tube walls is trans-ferred directly to the fuel support plate at the point where the'ower end fitting of the fuel element is supported vertically (Point 3).

The fuel support plates, acting as a relatively rigid diaphragm, transfer the in-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~gT F'iWTa L'age<

c>apl gc~

HMC)d LB Bkae QH&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 in 13-tube direction)

Fi nite El ement Fi xed-Base Fi xed-Base Fi xed-Base Interior Section ll Lumped-Mass ll Lumped-Mass 2 Lumped-Mass Without Model klithout Model Mith Model With Module with Added 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 ltoCO North-South Direct on Oo00 2X Damping Ct cc D.Ot

~.CO EC 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

~J LJ Colo

'.IO r

CC LJ Cal loCO I ~ CO O.CO O.IO IoCO IO.CO FREOUKIICT

~ IIERTZ North-South Direction 2C Damping Oo CO I

ECA'o00 LJ 0 CO O.CO IJ A

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

Ill II II III III Ill

~

~ ~

~

III Ill II lll III 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 inches vertically and impacts the top of a fully loaded YDFSS module.

The dropped assembly comes to rest horizontally on top of the HOFSS.

A fuel assembly drops from 18 inches above the HDFSS, enters an empty storage position and falls to the bottom of the storage position.

A fuel assembly drops from 18 inches above the HDFSS and strikes a tube wall at an oblique angle.

A fuel assembly drops from 18 inches above the top of a fully loaded module and strikes the upper tie plates of 2,

3 or 4 fuel assemblies in s.orage.

A fuel assembly drops from 18 inches above the HOFSS, falls

,outside of the loaded

HDFSS,

'and'odges

adjacent, and parallel to 'an unpoisoned, occupied fuel storage position.

Effect on Reactivit Analysis indicates that localized tube damage or fuel support member damage will occur, but neutron absorber material will not be r emoved from its position between adjacent fuel assemblies.

A fuel assembly, resting horizontally atop the HDFSS does not increase the system reactivity because the reactivity assumes an infinite vertical length of fuel (no neutron leakage in the vertical dimension).

.. keff (0.90 Structural ana lysis indica tes that local-ized tube damage will occur. and one neuter absorber plate may be damaged.

A reactiv,~,

analysis of this case with the neutron absorber p'late between two fuel assemblies

.totally missing, shows that keff remains less than 0.90.

It is not possible for a fuel assembly drop of 18 inches to drive four stored assemblies through the bottom of the module.

Even so, the reactivity effect of this impossible event was calculated as a

limiting value.

An 18 inch section of fuel in four bundles in an unpoisoned squar array was found to have a

k ff approximatel equal to that of the system.

There would be no increase in the overall. reactivity.

k ff 0.90.

This case was analyzed for normal handling conditions.

~

k ff < 0.90.

forces to the module base assembly plate (Point 4).

The forces are carried,in the-module base assembly (Point 5) until they are ultimately transferred to the module pad and to the support pad and the poolslab (Point 6).

The path of the vertical forces induced by earthquake motions is somewhat more complicated.

Ultimately, the vertical forces caused by earthquake and. gravity,loads b'ecome axial forces in.the. module pads.

The critical location for the-compression forces frcrt'a'the module pads 'is in,the long castings and tubes'irectly above the module 'pads;-'or stress analysis

'urpos'e, these compressive forces are considered to be resisted by four fuel 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 thermal stresses need not be considered.

Nevertheless, it was'ound that under the cooling-water flow conditions specified for the design, the heat rise in the storage tube wall due to gamma heat is less than 5 Fat the top of the module between tubes and negligible temperature difference at the bottom of the module.

Since the module is free to expand laterally at-

'he top of the 'module the thermal stress can be ignored.

The maximum water temperature rise from. the. bottom to the top of a storage tube is about 16oF.

Since the module is free to expand vertically, no thermal stress is generated.

Fuel assembly drop accidents were analyzed.

The results are summarized in Table 2-4.

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

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

l.

ASME Boiler and Pressure Vessel Code Section III, Subsection NF.

2.

USNRC, Reg.

Guide 1.92, Combining Modal Responses and Spatial Components in Seismic'esponse Analysis.

3.

Final Safety Analysis Report, Seismic Design Criteria.

OBE Operating Basis Earthquake SSE Safe Shutdown Earthquake 4.

Seismic Analysis of the High Density Fuel Storage System Browns 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

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 Appendix XVII, ASME,Section III, because of its appliability to these li'ght-gage tubes.

April 1978

0

0 3 ~

Materials 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..

These plates are sandwiched between the"inner and outer wall of the storage tube and are not subject to dislocation, deterioration or removal,,

deliberate or'nadvertent.

The inner and outer walls of the storage tube'are welded together at each end, thereby isolating the Boral" plates from direct contact with Spent Fuel Pool,(SFP) water.

At normal pool water operating temperature there is no significant deterioration or corrosion of stainless steel or Boral.

Specifications were developed specifically for the High Density Fuel Storage System which impose requirements to implement and follow accepted and proven industry standards during the design, procurement, fabrication, installation and testing of the storage system.

Periodic audits of the various facilities and practices are performed by certi-fied quality assurance personnel to ensure that these QA/QC r'equirements are being met.

All welding and nondestructive examinat'ion (NDE) is done in accordance with ASME Boiler '.6 Pressure Vessel Code and the American Society for Nondestructive Testing (ASNT) requirements.

St'orage module components are assembled and welded in special fixtures to maintain a high degree of dimensional tolerance.

Each storage position is checked with full length gauges to assure proper clearance between stored 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 the manufacturer's facility.

Samples of each Boral sheet are analyzed to determ'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 exceeds specification requirements.

Analyses are also performed to establish the correlation between the B

content and the thickness of the Boral sample.

The Boral sheets are dimensionally inspected and the thickness data are statistically analyzed to verify the sheet meets the'inimum thickness requirement over its entire area at a 95X confidence level.

These thickness data are also compared with the correlation data to provide additional assurance that the B

content meets or exceeds specification requirements.

Before each piece of Boral is inserted into a tube assembly it is verified that each inspection has been successfully performed.

Product of Brooks

& Perkins, Enc. consisting of a layer of B4C-Al matrix bonded between two layers of aluminum.

April 1978

- 21 The pr'esence of the neutron absorber material in the fabricated fuel storage module will be verified at the reactor stoiage-pool site by use of a neutron source and neutron detectors.

There will be a permanent record of all test results that will provide a comparison between 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 be accepted unless measurements'ndicate that'ive or more Boral sheets are not present.

Boral has corrosion resistant properties, similar to standard aluminum sheet.

Corrosion data and industrial experience confirm that aluminum and Boral have acceptable corrosion resistant properties for the proposed application.

Although experience indicates that it is un-necessary, an inservice test program will be conducted, consisting of periodically removing and examining, samples of Boral plate which have been 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 result of the planned modification to the spent fuel storage capacity (see section 3 '

of the environmental assessment).

0 USNRC Safety'valuation for Yankee Rove, dated 12/29/76 page 4, Structural and 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 pool I

1I.

Preparation - Equipment A.

Uncrate modules for inspection l.

Use tilting fixture to turn crated. modules from horizontal to vertical or upright position 2.

Remove crating and visually inspect.

Record.

3.

Clean as required 4.

Attach liftingtool assembly and transfer module to refuel floor B.

Verification of neutron adsorber material existence l.

Disconnect/remove lifting assembly 2.

Visually inspect each tube of module for demarcation lines showing 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 into position 2.

Align and shim support pad 3.

Remove handling fixture 4.

Repeat steps 1 through 3 until desired number of pads are in place B.

Installation of modules 1.

Attach liftassembly to module

2. 'ower into position and align module (No movement of materials over stored fuel is permitted by the technical specifications.)

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 pools will be accomplished by using the reactor building crane.

The redundant reactor building crane is described in BFNP FSAR section 12.2.2.5.

Nuclear The following assumptions were used in the analysis of the nuclear CritiCality of the 8/8'telil.

(a) 8 x 8 BWR fuel configuration (b)

(c)

(d)

{e)

Maximum BWR fuel bundle multipljcation factor (k

) of 1.35 in standard core. geometry at 20 C.

The use of a%aximum fuel k

as a criticality base eliminates the multiplicity of U-935 enrichment and burnable poison combinations and clarifies the exact condition considered.

Storage space pitch of 6.563 in.

Boron

{

B) equivalent

$o a )omogeneous areal concentration of 0.01'3 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 that separate fuel bundles.

The criticality analysis calculations were performed with the MERIT computer program, a Monte Carlo program which solves the neutron transport equation as an eigenvalue or a fixed source problem including the effects of neutron shielding.

This program is especially written for the analysis of fuel lattices in thermal nuclear reactors.

A geometry of 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-IV library tapes.

The qua) ification of the MERIT program rests upon extensive qualification studies including Cross Section Evaluation Work Group (CSEWG) thermal reactor 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 with alternate calculational methods.

Boron was used as solute in the moderator in the B&W UO~ criticals, and as a solid control curtain in the Jersey Central experimeAts.

The MERIT qualification program has established a bias of.005 +.002 (1 a )

6 k with respect to the above critical experiments.

Therefore, MERIT underpredicts k ff by approximately 0.5 percent 6k.

d.C.II

.EN i-IVBlMk1Hi ilF11 Spectrum Three Dimensional Monte Carlo Models, ANS Meeting Nov.

1977

25 The storage space (cell) infinite multiplication factor (k

) was calculated for the high density fuel storage system as deflated by the assumptions above and the exact geometry specifications.

Table 5-1 summarizes the results of the k calculations.

The maximum k of a storage cell occurs at 20 C witII the fuel bundles centered and IIo flow channels present.

Any variation such as increasing the cell pitch, eccentric bundle positionjng, reducing moderator density, and incr easing the temperature to 65 C decreases the k

Table 5-2 shows the maximum k of the storage cell broken down in% contributing bias and uncerIIainty values.

Figure 5-1 demonstrates the decrease in k with decreasing moderator density.

Since the cell is under moderated the optimum k occurs at 1.0g/cc.

The design of the HDFSS has alternating spaces on Ehe periphery of each module fabricated with unpoisoned closure plates.

The unpoisoned locations are also directly opposite each other on adjacent modules.

The effect of the partially unpoisoned storage locations is small and insensitive to the inter-module water gap as shown in Table 5-3.

The maximum module k

occurs at the minimum possible water gap {1.244") and is less than that of an infinite array of storage cells with no water gap.

The calculational model used to r epresent an infinite ar ray of modules is shown in Figure 5-2.

Each storage cell in Figure 5-2 is a simplified cell (Figure 5-3) used to reduce the required Honte Carlo input.

The final module k

includes the HERIT bias and uncertainty.

For the single or simplified cell calculation, no geometry bias would appear due to the fact that MERIT allows an exact representation of the geometry.

In conclusion, the HDFSS has a k

<.95 at a.95% confidence for all conditions analysed.

The design of the high density spent fuel storage system has spaces along the periphery of the storage pool for storage of defective fuel and control 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 pool storage 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 of a dropped module.

Since the new modules will not be transferred over the top of either new or spent fuel and will be transferred along a predetermined path, this type accident could result in only damage to noncritical component equipment.

26 TABLE 5 SINGLE CELL HIGH DENSITY FUEL STORAGE CRITICALITY RESULTS CASE DESCRIPTION Nominal Rack Dimensions ** With Flow Channel 9 20 C

k

(

2a)"

.8668 +.0075 Nominal Rack Dimensions Without Flow Channel 9 20 C

.8674 +

.0086 Same as Case 2 except 9 65 C

.8561

+.0084 Increased Pi)ch Without flow Channel 9 20 C

.8364 +

.0106 5

Same as Case 2 but with Eccentric Bundle Position

.8276 +

.0123

~includes MERIT Program Bias and Uncertainty at 9BB Confidence Level 00

• • 6.563" Pitch With Nominal Material Thicknessess

0

27 TABLE 5-2 BIAS & 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.

28 Figure 5-1 Cell k

Versus Moderator Density X

~J VIT IO I

~ q

.90

.80 4

~

+

.60

.50 4 ~ y r.w o

~ ~

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 uncertainties 00 at a 955'confidence level.

30 UNPOI SONED A = 1/2 Gap Between Inside Walls of Opposing Fuel Storage Location FIGURE 5-2 HDFSS MODULE 1/8 MODULE ARRAY FOR NUCLEAR CALCULATIONS

f J

31 H 0 2

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 are implied above.

Morf specific relationships are as follows:

a ~

b.

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

00 Stainless steel'hickness

- Neutron absorption by the two layers of stain ess stee comprssing the storage tube was included in the criticality calculations using the nominal thicknesses.

The sensitivity of stainless steel within the limits of the thickness tolerances is known but is not significant.

c.

Water densit

- Figure 5-1 shows the variation of koo with moderator water ) density.

The effect on reactivity of an accidental fuel assembly drop onto or adjacent 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 for any 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 an annual cycle basis.

The high density fuel storage system heat load for 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) from reactor shutdown to residence in t'e fuel Rtorspe pool.

A technical feasibility study on the use of an 18-month cycle is being conducted for Browns Ferry units 1 and 2. If these results are favorable, TVA will evaluate 18-month cycles as a planning basis for all Browns Ferry 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 was calculated.

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 shown as 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 batches for the alternative discharge cycles.

Discharges at the annual cycle rate will fi'll the high density system

{3471 assemblies),

less reserve for one full core (764 assemblies),

in thirteen cycles (years).

The thirteen cycles are shown on Figure 6-1.

Similarly, the 18-month cycle will fill the system in ten cycles

{15 years) as shown on Figure 6-2.

The maximum heat load was computed by assuming that the total reactor core is discharged just before a scheduled refueling date.

The total core discharge filled the last available spaces in the pool.

Core exposures at the time of shutdown were in the worst case conditions as given below:

Annual cycle 204 assemblies 204 assemblies 356 assemblies 1 year exposure - 8,400 NWD/NTU 2 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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36 18-month cycle 272 assemblies 492 assemblies 1.5 year exposure - .12,600 NWD/MTV 26,000 MWD/MTU It was further assumed that sixteen days cooling time was required to unload all of the fuel (5 days preparation and ll days unloading). Heat load results are shown in Figure 6-3. From reactor shutdown to the end of the discharge period, the heat. load decreased to approximately 28 million BTV/HR. Water flow through the passages in the storage module and support system w'ill be adequate to maintain cooling with the heat transfer systems available. With the incoming water to the fuel s~tora e pool at 125 F, the maximum fuel cladding temperature will be 165op, The maximum water temperature associated with the hottest fuel bundle wiU be 144'P. Those temperatures are low relative to structural integrity or corrosion limiting temperatur es of the structur al components of the storage system and fuel. Continuing efficiency of the exchange of heat from the spent fuel to the pool water depends on the convection flow of water through the storage location and flow channel encompassing a fuel bundle. The floe-like crud that adheres to the surfaces of the spent fuel bundle was studied to determine whether it is a potential mechanism for blocking flow through the Channel. It WaS fOund that the floe iS eXtremely fine, SuCh that pieCeS that spall off of the aggregate are not disposed to settle, but may flow upward with Che convection current. Additionally, the floe"is so fine that some of it will pass through conventional laboratory filter papers. Growth of crud in fuel storage conditions has not been observed in commercial facilities. The potential for channel plugging by sedimentation or by blockage of flow passages is negligible. The high density fuel storage system and the BWR fuel to be stored in it are not fabricated such Chat significant quantities of air or other gas can be entrapped creating an area of reduced effective moderator density..But eveo 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 this it is apparent that there is no possibility of boiling in the pool, thereby eliminating steam formation as a source of variations in moderator density. Two loss of cooling cases have been analyzed to indicate the relative temperature inertia of the expanded pool compared to its existing capacity. The first case is that of the full core discharge with the pool full of fuel discharged under the normal schedule except for the slots reserved for the full core unload. The worst condition, represented by the fuel exposures of the 18-month cycle, is used in this case. Cooling is lost at the completion of the discharge (sixteen days) with the pool temperature at 150'F. For the existing capacity, Case 2 assumes that one discharge of 272 bundles is already in the pool. At the time of the next discharge, the total core is discharged making a total of 1036 bundles (existing capacity is 1080). The exposure of the full core and initial pool temperatures are the same as those in (use 1. The calculational method used allows for evaporation from the pool surface, but not conduction through the pool walls. Results of the calculations are plotted 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 in parallel with the fuel pool cooling system and acts as a seismic Category I backup. No modifications are planned to be made to the existing pool cooling system because calculations have proven that the existing system has the capability of handling the expected heat load. The fuel pool cooling system is a redundant system consisting of two heat exchangers and two pumps with separate power supplies. In the unlikely event that both of these are inoperative or the heat load exceeds the capacity of 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 by the spent fuel pool cooling system and RHR, system parameters monitored in the main control room. Additionally, the temperature of the spent fuel pool water is recorded on a scheduled basis as required by technical specification. Should cooling be lost for an extended period of time such that the temperature of the pool became elevated sufficiently to cause evaporation of the pool

water, makeup to the pool could be supplied by means of a fire hose.

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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 Refueling Date Number of Fuel Assemblies Dischar ed Cumulative Number of Fuel Assemblies in SFP Sept. 1977 Sept. 1978 Sept. 1979 Sept. 1980 Sept. 1981 Sept. 1982 Sept. 1983 Sept. 1984 Sept. 1985 Sept. 1986 Sept. 1987 Sept. 1988 Sept. 1989 Sept. 1990 Sept. 1991 Sept. 1992 Sept. 1993 168 220 196 196 204 200 200 200 200 200 200 200 200 200 200 200 200 168 388 584 780 984 1184 1384 1584 1784 1984 2184 2384 2584 2784 2984 {Storage Limit)... 3089 (Storage Limit)... 3089

Brogans Fcr refueling Date March 1970 March 1979 March 1980 March 1901 March 1982 March 1983 March 1984 March 1985 March 1986 March 1987 March 1988 Narc+ 1989 March 1990 March 1991 March 1992 March 1993 March 1994 Number of Fuel 'Assemblies Dischar ec1 160 220 196 196 2o4 200 200 200 200 200 200 200 200 200 200 200 200 Cumulative Number of Fuel Assemblies in SFP 160 380 504 78o 984 1184 1384 1584 1784 1984 2184 2384 2584 2784 2984 3089 (Storage Limit) 3009 (Storage Limit) Browns Fer 3 Refueling Date Sept. 1978 Sept. 1979 Sept. 1980 Sept. 1981 Sept. 1982 Sept. 1903 Sept. 1984 Sept. 1905 Sept. 1986 Sept. 1907 Sept. 1988 Sept. 1989 Sept. 1990 Sept. 1991 Sept. 1992 Number of Fuel Assemblies Dischar ed 2o0 2o0 108 ,200 200 200 200 200 200 200 200 200 200 200 200 Cumulative Number of Fuel Assemblies in SPP 2o8 416 6o4 804 loo4 12o4 1404 1604 10o4 2004 22o4 2404 2604 2707 (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 18 months'ntervals. The existing capacity is 1080 bundles; 5 annual discharges or 4 discharges at 18 months. The heat load increment of the expanded 7 fuel 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, if this 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 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 will be 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 will be consumed by the proposed modification together with estimated annual domestic consumption are indicated below. k Material 304 Stainless Steel Boron Carbide Aluminum Browns Ferry Modification uantit 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. 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 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.

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 CONCLUSION

S 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

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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.

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