.

A -).

<

HEDL-TME 75-12UC-79b

COMPILATION OF DATA

AND DESCRIPTIONS

FOR UNITED STATES AND FOREIGN

LIQUID METAL

FAST BREEDER REACTORS,

J

l

E. R. Appleby

August, 1975.

Hanford Engineering Development Laboratoryc;erated By the for the United States

Wesfinghouse A $stsidiary of inergy Research and,, ,,,, p ,,3, g i,,,,,, o,,,,,,,,,,,,,,,,,,,,,,n

Hanford Company corporation contract u at 45 nmos

820517o076

CLINCH RIVER BREEDER REACTOR.

XVII.(CRBR)

LOCATION: Clinch RiverOak Ridge, Tennessee(TVA site)

OPERATION: Construction1976-1980 target

The objective of the Clinch River Breeder Reactor (CRBR, project is|>

breederthe design and construction of a 350 MWe demonstration fast

;

A joint proposal for the design, construction, and operationlley

of CRBR by Commonwealth Edison Co. of Chicago and Tennessee Vareactor.

In March 1972, twoAuthority was accepted by the AEC in January 1972. al. Thenon-profit groups were formed to implement the CWE-TVA proposd

Project Management Corporation (PMC) was to provide centralizeThe Breedermanagement of the Project and the demonstration program.

!

ii ts

Reactor Corporation (BRC,) served as liaison between project part c panBRC represents 350 utilities which have

and the utility industry. In November 1972, AEC and PMC announcedcontributed to the project. ime

the selection of Westinghouse Advan ed Reactors Division as pri the .

contractor, with responsibility for: designing and supply ngI l igNuclear Steam Supply System, and assisting PMC in overall p ann n .1972,

Burns and Roe was retained as architect-engineer in DecemberContractwith Holmes and Narver, Inc. assisting in selected phases.14,1973, andbetween Westinghouse and PMC was signed on Novemberh logy

the contract between AEC and Westinghouse for supporting tec noGeneral Electric Co. Breeder Reactor30, 1974.; tinghouse.was signed January

Department and Atomics International are subcontractors to WesThe schedule announced by PMC in January 1974 was for filing<

i rmitthe environmental report by December, 1974, construct on pe1976

expected about 14 months later, and a construction period fromThe first application was rejected because of lack of

detailed inforrstion on site meteorology, seismology, and aquaticto 1980.

environment.

XVII-I

_

R, .'

--...

Under a realignment of industry-government relationships in theLiquid Metal Fast Breeder Reactor (LMFBR) program in 1975, theEnergy Research and Development Administration (ERDA) will now

assume direct management of the CRBR Project, with PMC assuming anadvisory role along with the TVA group and a three member ProjectSteering Committee. '

The Nuclear Regulatory Commission docketed the Clinch Riverapplication in April 1975. Debates over the requirement for A.core-catcher are still progressing.

,

I..

.

XVII-2

.

CRBR XVII* -

INDEX

Pagg

XVII-5REFERENCES

XVII-7REACTOR PARAMETERS

SECTIONSXVII.A-1

A. CORE AND BLANKETXVII.B-1

B. CORE SUPPORT AND VESSEL INTERNALSXVII.C-1

C. REACTOR VESSELS AND SHIELDINGXVII.D-1

D. CONTROL ELEMENTSXVII.E-1

E. HEAT TRANSFER SYSTEMS ,

XVII.F-1S0DIUM PURIFICATION AND INSTRUMENTATIONF.

XVII.G-1G. COVER GAS AND AUXILIARY SYSTEMS,

XVII.H-1H. STEAM GENERATORS

.

XVII.I-lREACTOR INSTRUMENTATION AND CONTROL1.

XVII.J-l'

J. FUEL HANDLING,! XVII.K-1

K. CONTAINMENT /

>

I

XVII-3

:.

-.- - -.

I-

f3

,

CRBR XVII

REFERENCES

Clinch River Breeder Reactor Project, Proc. Breeder Reactor Corporation1.Jan.1974 Information SessionsPMC-04-01 (1974) (CONF-740116)

Clinch River Breeder Reactor ProjectNuclear Engineering International Vol. 19, #221 (Oct. 1974)} 2.

'- LMFBR Fuels and Materials Devehpment

,' G. W. Cunningham, p. 840-842, - Wall chart, p. 843| - The Clinch River Breeder Reactor Project Nuclear Steam Supply SystemI

I W. M. Jacobi, p. 846-850- Environmental Aspects of the CRBRP

J. H. Wright, p. 851-853- Safety of the Clinch River Breeder Reactor Project* J. Graham, et. al. p. 854-856

- Steam Generator for the United States LMFBR Plant857-859R. E. Skardahl, F.E. Tippets , p..

- CRBRP Reactor RefuelingK. W. Foster, p. 860-862

- Clinch River Breeder Reactor Plant and Balance-of-Plant DesignS. McPhersor), p. 863-864

Safety Related Criteria and Design Features of the Clinch River3.Breeder Reactor Plant.L. E. StrawbridgeProc. Fast Reactor Safety (Meetint, ANS, April 1974CONF-740401-P1, p. 72-92 1974)

ClinchRiverBreederReacterProfect,1974TechnicalProgressReport4.U.S. GPO 1975

Proposed Reference Design for the Clinch River Breeder Reactor PlantProc. Breeder Reactor Corp., Oct.1974 Information Session5.

PMC-74-02CONF-741087

XVII-5

A. ,a -- .

!I l.

D. |

CRBR XVII !.

:'

GENERAL PARAMETERS,

Table - 1 I

LoopType

3 (+ Overflow Heat Removal System)No. Loops

SodiumCoolant

730*FIl Core inlet temp.Il 995'F6

Core outlet tenp.

D UowardFlow directiong

Mixed oxide) Oriver fuel2 oxide}

Blanket fuel6.2 ft.

f Core diameter! 2400t

Core volume>

2Enrichment zones

350/975Output MWe/MWt

1.2Breeding ratio

Max. can temperature .1215'F

14.5 kw/ft. max. (7 kw/ft. ave)Linear po-sr

Doppler (-Th).0073 (Beginning of first cycle)

198Driver elements ~

,

e

I 150Rad. blanket elements

15/4 Primary / secondaryIn-core control elements

NoneIn-vessel storage

80,000 peakGoal Burnup

.

.

*

XVII-7'i

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

.

CRBR XVII

Table - 2CORE AND BLANKET

__

Driver Zone Axial Blanket Radial BlanketEuel materialUO2-Pu02 U02 U02

__

Form Pellet __

Pellet PelletStoichiometry

Substoichiometric

Enrichment Pu(U + Pu) .- ,!Depleted Depletedinner 22%

__

_32%_ !Outer

iRod cladding materialStainless steel 316 ;

_ Stainless steel e_

od 0.23 in. 0.23 in. 0.52 in.Wall thickness 0.015 in. 0.015 in. 0.015 in.

{Column length 36 in. 14 in. x 2 64 in. ,1'

Gas plenum length 48 in. (top)36 in. (top)

Vented-non-vented No

Assembly materialStainless steel I*;;; Stainless steel iShape aHexagonal ii

, Hexagonal 1

Across flats 4-1/2 in. | t.* *

4-1/2 in.Overall length 180 in. h h~ 180 in.

B_ Pins /assenbly 217 a *5 61

-

EPin spacing Spiral wire.

.

' g 'E Spiral wire'

Pin pitch1.26 P/D^

%i #Assembly pitch 4.76 in.i |!

*Assemblies in core 198

150Fuel .325

!!

;

XVII-8

. .. .

. .. . _ _ .

.

CRBR XVII.A.*

XVII.A. CORE AND BLANKET

CRBR fuel assemblies are similar to rings and the core barrel. Overallthe Fast Flux Test Facility (FFTF) lengths of all reactor assemblies and

the reactor vessel have been shortened.assemblies. The driver core consists of inlet and 3utlet nozzles of the assemblies198 assemblies in a triangular pitch of have been modified to improve handling4.76 in. Feed enrichments are 22% and location in the core lattice. Cnolantpu/U+Pu for the inner core zone and 32% flow distributions in the ducts and rodfor the outer zone for the equilibrium bundles have been improved by reducing thecycle. First core enrichments are 18.7%and 27.1%. Surrounding the core are 150 space between the outer rods or pins.and

rcdial blanket assemblies. The reflector the hexagonal can. Smaller spacer wiresare used on the outer row rods, whileand restraint assemblies quoted in the

General Parameters have been replaced by maintaining the same spacings. (Fig.

174 rcmovable radial shield assemblies toXVII.A-1,A-2) (Ref. 4)

protect the two core restraint former

|;

*

l .

i

k

I

i:

!l

,

:s

i

*

;

E

XVII.A-1

!

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y_

. . _ .

u _ m . _m

?

CRBR XVII.A.

1 2 3' 4 5 5 6 7' 8 9 10[ - __ e N

/M~~~^~'~xr/

oc Pc

@ 9 c,

(| pc ic ic ac se ac ec oc oc no as as as as asac

h h ec

c ac ac

t -.

1.,

-

_

bQs.. . ,..53 Component Name Quan.

1. 77.50 R 6. FUEL TRANSFER / STORAGE IC INNER ZONE FUEL ASSEMBLY 1082. 75.50 R 8.625 0.D., 7.981 !.D.OC OUTER ZONE FUEL ASSEMBLY 903. 75.00 R (4 PLACES) R8 RADIAL BLANrET ASSEMBLY 1504 70.00 R 7. 5.00 FIXED SHIELDING R5 RADIAL SHIELD ASSEMBLY 3245. SURVEILLANCE 8. .50 GAPPC PRIPAR) SYSTEM CONTROL ASSEMBLY 15SPECIMENS 9. 2.00 CORE BARRELSC SECONDARY SYSTEM CONTROL ASSEMBLY 410. 85.70 R

Figure XV11.A-1 Reactor Cross Section (Ref. 4)

XVII.A-2

_ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . . _ _ _ _ . _ _ . . _ _ _ _ _ _ _ _

i

l

,

CRBR XVII.A.O6

t'o4

e

s.. . ..,d 3. . ... . , .' ..

I

.1 .j I f 6.. P e .. .

88 . * l| ,'*P,

)

i ... . .!

. s..

e u.e& 'f< * = '

l

i'

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I.7'' ) **-- *,';d ****,,

!d I . . .,

,

la. f .. . .. . .-.~.ii....

e

i

i

', Figure XVII.A-2a Core Fuel Assembly (Ref.1)

|<

j .-

.

|

||

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... ... Q. ... . <e 3.......

f. .c .

|.

h ', t ,.e ...., .,a ,. . . . .

g,,.-

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i ,. '........i';< .

.l

1 . .. .... i'

.i....e ~\...*

i

$ 1 f......... ... gI~,

.- . .

,,

$1

i

Figure XVII.A-2b Radial Blanket Assembly.*

XVII.A-3

t1v

EYt

4s,'h,

~ '~ - ___ - _______________ _ _ _ _ _ _ - - - - - - - _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _

CRBR XVII.B.

XVII.B. CORE SUPPORT AND VESSEL INTERNALSThe skirt of the bottom support plate

The lower internal structure consists attaches to the reactor vessel wall atof the core support plate and core support a ring forging. The core barrel is askirt, core barrel, horizontal baffles, welded structure of 2 in. thick plate,core restraint fortner rings, fixed radial 151 in. dia. It is supported on theshielding, and inlet and by-pass flow support plate and is external to themodules. radial shielding. The core barrel rings

provide lateral restraint for the reactorThe core support structure is assemblies. Constraint fortner ringscomposed of a 24 in. thick core support apply against load pads, above the coreplate with holes into which if ners are The upper internals structurezone.inserted. The inlet nozzles fit into the utilizes transverse plates connectingliners. Each module holds the nozzle four jack-operated lift columns used tofor a group of reactor assembifes, and raise the structure during refueling,most modules have flow orifices to assist (Fig. XVII.B-1, B-2, B-3)flow distribution. The orifices alsoestablish pressure zones upon which Outlet modules above the assemblyhydraulic balance of the fuel, control, outlet nozzles guide the flow into theand radial blanket assemblies is based. outlet plenum. Each outlet moduleWith the hydraulic balance arrangement, collects the flow from a group of reactorthe upper core plate is not required. assemblies. (Fig. XVII.B-4)

.

i M 5"as*

E'',*",5,}f" 0m Q-

..s*. m . n ,. . s.

... . . .

* . .' y'' ** *

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- .

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fou LYT .= ~

.h aa%''"o. _ .

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TUut W CTT79' SCAL towtp fiEAL DET84-TLat

CRBR Core Support Structure (Ref. 5)Figure XVII.B-1

XVII.B-1

... . . _,

t..

CRBR XVII.B.

Support modules*,

* =

f fff8fffffffff' *

d).

,

i ! l

N , J J J J k'P|P|hiFi F F F FM: =

n= n _ ",_ ; = <= '

0 0 0 0 0 0 0 0

% gin;**"yA y -

-

/ f

NBypass flort ~~ ,,

*Main flow -----

|

)

l

|1

Figure XVII.B-2 CRBR Modular Core Support System Assembly Flow Paths (Ref. 1)

XVII.B-2

.. . - .

_

.

CRBR XVII.B.

'RelvelingCrapple Socket

I fsf p Care Formera

s* Q

167SD ?.ead Pads

,ns..e eore Fer.e,

[ - -sn.eiding

8:: -

M|

;i --

1

dC.,e piii

Figure XVII.B-3 CRBR Core Restraint (Ref. 1)

*.

I

-Centrol ed Drese

[~w-

-Flowenster Conduit|

' Nm.Thersnoceuple Conduits

i I | !

Upper Core SupportStructure

LJ CThermal liner

i l'

jFlow Collector

rw~

Core Assembly SupportGrid

_ ..-t

.

|L 1|

|sI.!! Figure XVII.8-4 CRBR Upper Package Outlet Module (Ref.1)..

Is|' XVII.B-3

||

_-

.

CRBR XVII.C.

XVII.C. REACTOR VESSELS AND SHIELDING.

The reactor vessel is a vertical The guard vessel provides secondarycylindrical shell with a torospherical containment and limits total leakage inbottom head and an upper flange. Inside the event of a leak from the primary vesseldiameter is 243 in., overall length 658.35 or piping. The cylinder has an id ofin. Material is stainless steel, with 262.5 in. and is 570.5 in. high. A skirtalloy steel used for the flange and bolted to the floor of the reactor cavity

stationary outer ring, and a transition supports the guard vessel. Leakage sensorsshell of Inconel between the low alloy and in-service inspection equipment can beand the stainless steel. The vessel located in the space between the twoflange is bolted to the stationary outer vessels. The inlet piping downcomer isring, and the ring is supported by the protected by an extension of the guardconcrete ledge of the reactor cave. vessel. (Fig. XVII.C-4)(Fig. XVII.C-1)

The reactor vessel cavity is sealedThe closure head assembly is a and filled with an inert gas. Below the

21-1/2 in. thick forging of low alloy cavity there is an area which will housesteel. It has radiation shielding, thermal an ex-vessel core retention device ifinsulation plates, and a suspended vortex- required.suppressor plate. The head is composed of .three independent rotating plugs, contained Individual cells hcuse the beatone inside the other. The ex-vessel transport system loops. All primarytransfer machine is located in the largest system components are surrounded by guardplug, and the in-vessel transfer machine vessels. (Ref.4)in the smallest plug. (Fig. XVII.C-2,C-3)

*.

I

i

1

||

|

XVII.C-1

!

|

._

'-

.

CRBR XVII.C.

E1

s .-.

sit

2

5 7 N ,I| ESS5:hi

s[ !3

s:. P :-

; 'l I'

4,

'7 . s

i ,, ,. ,

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x y % is,. - .. 4

% $-.' 'I

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8>s !,; .

-

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X_ -6 "? ? , ' , '

9

10:.

e ,e-

,een

I et*

.t|

g ||L

1. PLUG RISER 6. VESSEL FLI,NGE2. REACTOR CAVITY SEAL 7. CONCRETE LEDGE3.CLOSURE HEAD VESSEL HOLDDOWN t10LT8. INCONEL 600 TRANSITION4 STATIONARY OUTER RING 9. HEAD TO VESSEL BOLT5. O!dEGA SEAL

10. VESSEL WALL

Figure XVil.C-1 CRBR Vessel Support System (Ref. 4)

XVII.C-2

4

.

CRBR XVII..C.

):i::

c' v=,| q=y )_ic

_. . .s ,||

j_ _. . _ .

-

d- 1 * #2y-- &_-:_- m m

- ,.- :3

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,

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fib [1,

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-

17

gg;%$-" m !8

l ) UU (/i ,9

b: I a t:1 'u '_ 10''

t t,

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d |N_

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4 /'! 12 .

~h k ;13.'

::5:.'?'

, .

. * g. X . ' ;14.. - :.

1. COVER GAS 8. RADIAL SHIELD ASSEMBLIES

2. SODIUM LEVEL 9. FUEL TRANSFER AND STORAGE POSITIONS

3. VORTEX SUPPRESSOR PLATE 10. REACTOR CORE4. OUTLET PLENUM 11. CORE SUPPORT STRUCTURE

.

5. UPPER INTERNALS STRUCTURE 12. INLET MODULEl

6. OUTLET N0ZZLE 13. INLET N')ZILE'

| 7. CORE RESTRAINT FORMER RINGS 14. INLET PLENUM

Figure XVII.C-2 CRBR Reactor System (Ref. 4)

1,

XVII.C-3

l

:-

. ._. _. __ __-__ _ __

g_,...

CRBR XVII.C.

I 2 3 4 5 6 7 8 9e o

E C g

= ae a c

~

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l D~ ~ ~

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= = = = = =W

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,%

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-

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,

1EX-VESSEL TRANSFER MACHINE PORT5.

IN-VESSEL TRANSFER MACHINE PORT2. INTERMEDIATE ROTATING PLUG 6. SHIELDING3. CONTROL ROD DRIVE MECHANISH 7. SUPPRESSOR PLATE4. SMALL ROTATING PLUG 8,LARGE ROTATING PLUG

9. STATIONARY RING

Figure XVII.C-3 CRBR Closure Head Assemoly (Ref. 4)

XVII.C-4

_ - _- - ___ _- _____ _

.

CRBR XVII.C.

1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 18

_ . . . ,

N | | -

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0'.0 g'36r3

|, e=. > <

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J.-,.,

1. REACTOR CORE 10. CONTROL RODS2. INLET N0ZZLES (3) 11. HEAD' HEATING AND COOLING SYSTEM3. STATIONARY OUTER RING 12. CABLE HANDLING

| 4. PLUG DRIVE AND CONTROL 13. EVTM PORT! 5. SEAL HEAD ACCESS AREA 14. BEARINGS

6. SEALS 15. HEAD ACCESS AREA7. IVTM PORT 16. OUTLET N0ZZLES (3)8. ROTATING PLUG (3) 17. GUARD VESSEL9. THERMAL INSULATION AND SHIELDING 18. REACTOR CAVITY

Figure XVII.C-4 CRBR Reactor Enclosure (Ref. 4)

XVII.C-5,

.

- %str. "

.

CRBR XVII.D.

XVII.D. CONTROL ELEMENTS

There are fifteen primary and four are independent. Pin bundles consist ofsecondary control assemblies. Eachprimary system control rod has 37 absorber SS tubes containing B C pellets. The4

rods are wire-wrapped and include a gaspins. Secondary assemblies each have 19 plenum. The outer duct remains in theabsorber pins. The secondary system core. The inner duct is connected toutilizes a hydraulic scram assist thethe control rod drive and moves relativeprimary system a spring assist. Primary to the outer duct. (Fig. XVII.D-1)and secondary reactivity control systems (Ref. 1, 4)

.

'

lt

lb

,

l-| .

| .4

Handling Socket

m ly Duct'

/r;/

Moveable Control Red,' (Neutron Absorber,

'

" }' Pin Assembly)

'

Shielding Orifice Assembly- *

,,

MInlet Nozzle*

Figure XVII.0-I CRBR Control Assembly (Ref. 1)

XVII.D-1,

.

- -___ _ _ _ _ _ _ _ . _ _ _ _ _ _ _

C, 34 "y,] > .1.

s

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t. ' ,

- +

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_-_

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CRBR XVII.E.'

!~ '

.

w,..

' s- _, /,'. * *, s '

*.. ,

.

iXVII.E. HEAT TRANSFER $YSTEMS '% ,

,

.t

=

There are three'secer.dary indepen- % cicsures. The tilting. disc seat: . against ~

[ dent loops. The primary system includes a free tan 11ng seat. K4 dyihpot controlsa hot-leg pump, the sbolliside of an the rat of disc closure upon f bw reversal,

I intermediate her.t exch 4 9er, and a check and limits pressure surges'in the loop.q

valve. No isolation valves are used. A flanged access opening 1s provided atpiping between the reactor and-the primary th o top of.the valve. (Fig., XVII.E-3)

$ pump is 36 in., betwesn the pump and the ,

heat exchanger 24 in. All primary loop The reiference design provides for2

f' piping is either elevated or contained forced convection decay heat removal andwithin guard vessels, and there are guard heat reur.) val by natural circulation, under

i vessels around the pumps and the normal or off-normal conditions. Eleva-Intermediate Heat Exchanger (IHX) Reactor tion of piping and the provision of guard

fr vessel outlet nozzles are sl6 f t. below vessels assure continued det.ay heat

> the normal sodiun level in the reactor remov31, with primary pumps operating,r.t

J vessal. pony rutor ' speed. The steam generatorauxiliary heat remval . system functions

The intermediate syste;n includr:s a f by first venting steam to the atmosphere.cold leg pump, the tube side of the IM, in the. event of lots of| normal feedwater,i

and incorporates an expansion tank in steam is generated from water from a,

each loop.' The system is arrcnged to protected storage tank.. Steam from tneaccommodate " hockey stick" steam generator steam drum is naturally circulated to thenodules (evaporator and super heater). tube-side of an air-cooled condenser,

condensed, aad returned to the steam drum,s

preliminary design for the primary ~ A: major change made recently is the

d and secondary pumps is being,done by addition of a system to allow decay heatremoval withuut the use of the main coolants

- Byron Jackson and Westinghouse-EMD. . -

/ Vertical centrifugal units are proposed. loops and the steam generators. Thisj The primary pump operates in the hot leg system, the Overflow Heat Removal System

of the loop at 995'F. intent at the (OHRS), was created by adding a heatexchanger to the existing overflow loopy

: present time is to make the. primary and *,

secondary pumps hydraulically identical.y and adding two small air-blast heat(Fig. XVII.E-1) i evchangers. The OHRS is rated'at 5.6 MWt.-

i f (Fig. XVII.E-4) (Ref. 4)The design concept for the Ir.ter. ,-

' rediate Heat Exchanger is a straight tube All loop vaults are independent and

.

component, with a flexible downcomer, shielded from one another. All main? (Fig. XVII.C-2) (Ref. 5) components are supported by their flanges

at the operating floor level in the-|

The check valve concept under containment building. Vault atmosphere isy1 consideration uses a tilting disc housed nitrogen. (Ref. 5)M within a cylindrical body with conical end!

l;f,

!..

E

\

!

XVII.E-1l

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y yn

.

CRBR XVII.E.

10

I 2 3 4 5 6 7 8 91

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& j No.:Q-Ln':;*f::q

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XVII.F. SODIUM PURIFICATION ANDINSTRUMENTATION

The auxiliary liquid metal system the auxiliary coolant fluid systemprovides for receipt, storage, and consisting of an electromagnetic pump,purification cf liquid metal, and provides diffusion-type cold traps, heat exchangers,the capability for reactor sodium level a storage tank and an expansion tank.control, accommodates primary sodiumvolume changes, and provides cooling for Hydrogen and oxygen meters monitorcore components stored in the ex-vessel sodium after exit from each unit, thestorage tank. There are three 60 gpm cold bulk sodium in the intermediate loop coldtraps for purification, one active trap leg, and the argon cover gas in thein the primary sodium processing subsystem expansion tank. Each monitoring locationand one in each of the intennediate has an electromagnetic pump, flow-meter,sodium processing subsystem. Cold traps heat exchanger and heater.have NaK cooling, with heat rejected to

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XVII.G. COVER GAS AND AUXILIARY SYSTEMS

Liquid metal coolant cover gas is The auxiliary liquid metal system providesfor receipt, storage, handling andAn impurity monitoring and analysis purification of sodium and NaK. Theargon.

system semples plant liquid metal and rad-waste system collects, stores, monitors.argon systems. Equipment is cooled by anj

} auxiliary system using an organic fluid. and packages and transfers liquid and-

A recirculating gas cooling system provides solid radioactive waste materials,

cooling to inert cells and equipment. (Ref. 2)

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XVII.H. STEAM GENERATORS

There are three steam generator tank or the storage tank of a centrifugalloops, one for each intermediate loop. separator. There is a sodium leakThe steam generator module chosen for the detector at each rupture disk location.>

reference design consists of two evapor- (Ref. 4)ators and superheater per loop, utilizing

' the hockey-stick concept. Evaporators A sodium dump subsystem is providedand superheaters are identical and to remove sodium from an affected loopinterchangeable, except that inlet water of the intermediate sodium system inorifice inserts will be added to the case of a sodium / water reaction, beforeevaporators. Overall height of the the IHX has been excessively contaminated.

, evaporator or superheater is 65 ft. Drain lines located at low points includeshell od. 52 in. There are 757 heat pairs of dump valves in series. Drain

1 transfer tubes 0.625 in od with single lines route the sodium to a dump tank0.109 in. thick walls. Active tube located below the appropriate superhaater/

i length is 46 ft. Tube triangular pitch evaporator cell of the steam generatoris 1.22 in. Sodium flows in the shell building. (Ref. 4)

4 side and water / steam in the tubes. Ahydrogen diffusion tube detector was-

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detector. (Fig. XVII.H-1) removal system removes decay heat and/orsensible heat from the reactor and heatj The sodium / water reaction pressure transport system in the event the mair

a s unavaHaMe. R can be] relief subsystem protects the sodium sideactivated when desirable. The concept' of the evaporators and superheaters and

the intermediate loop by the use of inc rporates two subsystems; a long-rupture disks in the main sodium piping term system and a short-term system.

** # " * * "YS ** "**8 * "close to the superheater and evaporatorr

nozzles. There are three rupture disks steam dug to ne am@em 3mugh poweri' relief valves in the steam lines. Theper loop. Pressure-relief system piping !

routes the reaction products to a 1 ng-term system uses three protected.

a c e coMensers on de mof of beseparation tank. Gaseous products arevented to the atmosphere, liquid and | s eam genera r buMng to coMense

* *** "" * * *** " **solid wastes remain in the separation

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The overall plant control system is Fuel failure monitoring is similardesigned to provide load-following to the FFTF system, with three subsystems.capacity at a maximum of 3% per minute The cover 9as monitoring subsystem includes

1 over a load range from 40% to 100% of a charcoal chromatograph column, a.

full power. Basic control is from a germanium detector, a fixed-channelcentral supervisory control system, spectrometer, a multi-channel spectrometer-providing for either automatic load computer and printer, gas samplingdispatch or operator load control equipment and associated control panels.,

capability. Main areas of control are The delayed neutron monitoring systemthe reactor, primary and intermediate includes BF3 neutron detectors and the,

ij sodium flows, primary and intermediate shielding moderator assembly, adjacent

I sodium pumps, and feedwater flow. Two to each of the three primary loops. TheC out of three logic is maintained through- failed fuel location system includes gas

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]system. (Fig. XVII.1-1) (Ref. 5)

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XVII.J. FUEL HANDLING

The EVTM reference concept isA new core component is received asa fully assembled element in a single- essentially identical to the FFTF CLEM.assembly shipping container, with twelve Changes include the deletion of abnost

all in-vessel storage, requiring removalcontainers per shipment, and transferred of spent fuel during the same refuelingin the containers to the new fuel storage shutdown during which it is removed fromcell. It is removed from the container Decay heat removal capabilityinside the cell and stored in separate the core.thimbles. About a month before scheduled required is $20 kw. Alternate approaches

to cold-wall heat removal system arerefueling the assembly is removed from being studied, for exanple forced con-storage by a new component transfer vection of argon, although it is concludedmachine to the ex-vessel storage tank, a that the cold wall EVTM will be capable oflarge two tier sodium tank with 650 handli,ng fuel with 20 kw decay heat. Thepositions. Reactor port olugs are removedand the In-Vessel Transfer Machine (IVTM) height has been reduced from 61 ft. tois installed. The Ex-Vessel Transfer about 34 ft. (Fig. XVII.J-2)Machine (EVTM) moves a new fuel assembly A two-tier, rotatable storage rackfrom the ex-vessel tank while the IVTM was selected for the ex-vessel storageand rotating plugs remove a spent assembly tank, where 650 core assemblies can be storedfrom the core and place it in an in-vesseltransfer position. The EVTM places the in ten circular rows. A fuel transfer portnew fuel in another in-vessci transfer for each circular row is excessible to theposition and removes the spent fuel EVTM. Storage positions are cylindricalassembly, transferring it to the tank. tubes restrained and supported by a .

Tne IVTM and the rotating plugs install stainless steel grid in the rotatablethe new fuel in the core open lattice rack. Each tube can hold two coreposition. All fuel handled by the EVTM is component pots, one above the other. Therein a sodium-filled core component" pot, are 24 argon-filled positions where new

. Heat is dissipated from the machine by assemblies will be placed for radiant pre-forced air cooling on the outside of the heating before insnersion in sodium. Twocold wall. On completion of refueling independ nt and redundant heat removal

the IVTM is removed from the reactor. , systems are provided, each capable of re-moving the maximum design decay heat andSpent fuel decays for one hundred days

in the ex-vessel tank before being J holding sodium outlet temperatures below'

shipped off-site. Maximum use of FFTF f600*F. (Fig. XVII.J-3)technology has been used in design and Spent fuel is transferred from EVSTdevelopment of the equipment. storage, af ter a suitable decay period,;

The IVTM conceptual design employs by the EVTM to the Fuel Handling Cell.a rising-stem grapple drive. Seals and The pot containing the spent assembly is

lowered into a sodium-filled storage tankbushings, drive components, grapple in the floor of the cell. An in-cell cranealignment, and IVTM prototype programs will lower the assembly through a port in ,

are under way. (Fig. XVII.J-1) the cell floor into the spent fuelshipping cask. (Fig. XVII.J-4) (Ref.4)

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Figure XVII.J-1 CRBR In-Vessel Transfer Machine (Ref. 4)

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XVII.K. CONTAINMENT

The containment building concept is centerline. A 7 ft. thick hollow cylinderfor a cylindrical steel vessel embedded extending from the base mat to 14 ft.. ,

in concrete up to the operating floor, below the operating floor supports the !i

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The front cover carries an aerial view of the EBR ilsite at the National Reactor Testing Station, Idaho.

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EditorErle H. HillEditorial AssistantJudy Denny

The Technical Review is published quarterly bythe Project Management Corporation and distrib-uted to members of the Breeder Reactor Corpo-ration and to colleges and universities throughoutthe country. For more information contact:Breeder Reactor Corporation, P.O. Box U, OakRidge. Tennessee 37830. Telephone: (615)482 9661. ext. 476.

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