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COMMISSION ORNL- TM- 3229 COPY

NO.

DATE

November 19, 1970

FLUID DYNAMIC STUDIES OF THE MOLTEN-SALTREACTOREXPERIMENT (MSRE) CORE R. J. Kedl

ABSTRACT In the MSRE reactor vessel, fluid fuel is circulated at 1200 gpm down through an annular region and up through 1140 passages in the graphite core. The core design was based on preliminary tests in a one-fifth scale model, followed by detailed measurements with water solutions in a full-scale mockup of the reactor vessel and internals. This report describes the models, the testing, and the data from which velocity, pressure drop and flow patterns are deduced. It also describes how the measurements were extrapolated to molten salt at 1200'F in the actual reactor. The few observations possible in the reactor

were consistent

with

the predicted

behavior.

KEYWORDS:

MOLTEN-SALT REACTORS,CORES, DESIGN, DEVELOPMENT,FLUID-FLOW, mRE, REACTORVESSEL, FLOWMEASUREMENT,MODELS HOTICEThis

d ocument contains information of 0 preliminary nature and was prepared primarily for internal use at the Oak Ridge National Laboratory. It is subiect to revision or correction and therefore does not represent o final report.

*â&#x20AC;?

This report was prepared as an account of work sponsored by the United States Government. Neither the United States nor the United States Atomic Energy Commission, nor any of their employees, nor any of their contractors, subcontractors, or their employees, makes any warranty, express or implied, or assumes any legal liability or responsibility for the accuracy, completeness or usefulness of any information, apparatus, product or process disclosed, or represents

that

its

use

would

not

infringe

privately

owned

rights.

iii

CONTENTS \

ABSTRACT INTRODUCTION

DESCRIFTION OF MSRE CORE AND TEST PROGRAM

One-Fifth Scale Model

F u l l Scale Model

DESCRIPTION AND ANALYSIS OF TEST RESULTS Volute and Core Wall Cooling Annulus

9 9

Reactor Vessel Lower Head

Graphife Moderator Assenibly

Reactor Vessel Upper Head

-3

Miscellaneous Measurements

EXPERIENCE WITH THE MSRE

REFERENCES

LEGAL NOTICE This report was prepared as an account of work sponsored by the United States Government. Neither the United States nor the United States Atomic Energy Commission, nor m y of their employees, nor any of their contractors, subcontractors, or their employees, makes any warranty, express or implied, or assumes any legal liability or responsibility for the accuracy, completeness or usefulness of any information, apparatus, product or process disclosed. or represents that its use would not infringe privately owned rights.

INTRODUCTION The MSRE (Molten-Salt Reactor Experiment) i s a 7 . 3 MW fluid-fueled,

graphite-moderated,

single region nuclear reactor.

The f u e l consists of

uranium fluoride dissolved i n a mixture of l i t h i u m , beryllium and z i r conium fluorides.

A unique feature of t h i s reactor concept is t h a t t h e

power i s generated i n c i r c u l a t i n g f l u i d f u e l r a t h e r than stationary s o l i d f u e l elements.

The nominal operating temperature i s 1200'F.

A detailed

description of t h e reactor concept and i t s components i s available i n many sources, f o r instance References 1, 2 and 3.

A program w a s undertaken t o

determine t h e f l u i d dynamic and heat t r a n s f e r c h a r a c t e r i s t i c s of the core.

This report presents t h e r e s u l t s of t h a t e f f o r t .

Most of t h e experimental

r e s u l t s presented here were obtained i n the early 1960's.

This report was not

written, however, u n t i l a f t e r t h e MSRE' nuclear operations were terminated '\

i n Deceniber of 1969. DESCRIPTION OF E R E CORE AND TEST PROGRAM Figure 1 shows an isometric view of the E R E core.

The fuel e n t e r s

t h e reactor vessel at 1200 gpm through a constant f l o w area volute near t h e t o p of the c y l i n d r i c a l section.

gradient

Because of t h e variable pressure

avolute of t h i s type, o r i f i c e s a r e used t o obtain a uniform

angular f l a w distribution t o t h e core w a l l cooling annulus.

The f u e l then

s w i r l s down t h e core w a l l cooling annulus and i n t o t h e reactor vessel lower head. R a d i a l vanes are placed i n t h e lower head t o destroy t h e s w i r l generated by t h e volute.

The lower head then serves as a plenum t o d i r e c t t h e

f u e l uniformly t o t h e moderator region.

The moderator region i s composed

of long graphite core blocks, square i n cross section, and with grooves cut longitudinally i n t h e 4 faces,

When these s t r i n g e r s are assembled

v e r t i c a l l y together, t h e grooves form t h e f u e l passages.

Figure 2 shows

an isometric and a plan view of a small c l u s t e r o r core blocks. passing through t h e moder z

-2

head which serves as a c o l l e c t

After

t h e f u e l then goes i n t o t h e vessel upper plenum and d i r e c t s t h e fuel t o t h e o u t l e t

pipe. Each of these various regions of t h e core w i l l be described i n more d e t a i l i n the appropriate section of t h i s report.

c) ORNL-LR-DWG 6t097RtA

GRAPHITE SAMPLE ACCESS PORT

FLEXIBLE CONDUIT TO CONTROL ROD DRIVES

COOLING AIR LINES

ACCESS PORT COOLING JACKETS

FUEL OUTLET

REACTOR ACCESS PORT MALL GRAPHITE SAMPLES OLD-DOWN ROD OUTLET STRAINER

CORE ROD THIMBLES LARGE GRAPHITE SAMPLES CORE CENTERING GRID

CORE SUPPORT FLOW DISTRIBUTOR VOLUTE

GRAPHITE -MODE CORE BLOCKS FLOW DISTRIBUTION ORIFICES CORE WALL COOLING ANNUI REACTOR CORE CAN REACTOR VESSEL

ANTI-SWIRL VANES

FIGUFE 1.

MSRE REACTOR VESSEL

ORNL-LR-DWG

56874R

f MODERATOR STRINGERS SAMPLE PIECE

NOTE: NOT TOSCALE

FIGURE 2.

TYPICAL GRAPHITE CORE BLOCK ARRANGEMENT

4 The

The MSFE core development program was divided i n t o two phases.

first phase consisted of building a l / 5 l i n e a r l y scaled p l a s t i c model and t e s t i n g with water.

This was considered t o be a rough and preliminary

The second phase consisted of b d l d i n g a full

design checking device.

scale carbon s t e e l and aluminum model and t e s t i n g w i t h water.

The only

data presented i n t h i s report w i l l be from t h e f u l l scale model, however,

a brief description of the 1 / 5 scale model and t h e way it w a s used i s given i n the next section. Early concepts of t h e MSRE c a l l e d f o r a variable speed pump. It was planned t o operate t h e reactor at f l o w rates below t h e design f l o w of 1200 gpm.

The lowest flow rate was undefined but could have been as low

as 25% of design flow.

Later i n t h e design stage, t h i s reduced f l o w spec-

i f i c a t i o n w a s dropped and design flow r a t e was fixed at 1200 gpm. change occurred during t h e t e s t i n g of t h e full scale core model.

This

As a

result, data were taken at flow rates ranging from 1200 gprn t o 300 gpm, b u t t h e emphasis i n t h i s report i s on t h e 1200 g p m data. Late i n t h e 4

operating h i s t o r y of t h e MSFE, t h e fuel pump w a s connected t o a variable frequency u n i t , and the reactor w a s operated at reduced f l a w r a t e s . purpose of these s p e c i a l runs w a s t o study Xe-135 behavior.

The

The "worst

case" as far as lateral temperature gradients i n t h e core i s concerned, would be vhen it w a s operated a t 5.5 MW (thermal) at h a l f t h e design flow rate f o r a period of about 3 1/2 days.

Figure 3 i s a l i s t of reactor parameters and physical properties of i n t e r e s t i n t h i s study. One-FiFth Scale Model A s m a l l transparent p l a s t i c model of t h e MSRE core, l i n e a r l y scaled

dawn by a f a c t o r of 1/5, w a s b u i l t and t e s t e d .

The p a r t i c u l a r core com-

ponents simulated i n t h i s model were t h e i n l e t volute, flow d i s t r i b u t i o n o r i f i c e s , core w a l l cooling annulus, lower vessel head with s w i r l k i l l i n g vanes, moderator sppport and t h e moderator f u e l channels which w e r e simulated with a tube bundle.

A photograph of t h e model i s shown i n Figure 4.

The p a r t i c u l a r scale factor of l / 5 w a s chosen because a geometrically

similar model of t h a t s i z e t e s t e d with water at about 95OF, and at a flow

rate such t h a t i t s f l u i d v e l o c i t i e s a r e equal t o those of t h e r e a c t o r , w i l l have t h e same Reynolds Number as t h a t i n t h e actual reactor core when

DESIGN CONDITIONS FUEL SALT FLOW RATE REACTOR POWER FUEL INLET TEMPERATURE TO CORE FUEL OUTLET TEMPERATURE FROM CORE REACTOR OPERATING POWER

9Pm

1200 10 Mw t) 1175°F 1225°F 7.3 M w ( t ) ~

FUEL SALT COMPOSITION L i F BeF2 ZrF4 UF4 LIQUIDUS TEMPERATURE PROPERTIES AT 1200°F DENSITY SPECIFIC HEAT THERMAL CONDUCTIVITY VISCOSITY PRANDTL NUMBER = (0.47)(19)/(0.83) E

65.0 mole 29.1 mole 5.0 m o l e 0.9 mole 813°F

GRAPHITE GRADE POROSITY (ACCESSIBLE TO KEROSENE) WETTAB IL I T Y FUEL SALT ABSORPTION AT 1 5 0 p s i (CONFINED TO SURFACE) UNIRRADIATED

8.79 11.71 B t u / h r f t "F 0.139 B t u / l b OF 7.81 x 10-6/oF CGB 6.2%

0.2% IRRADIATED

117 l b s / f t 3 0.42 B t u / l b "F

117 l b s / f t 3

80 B t u / h r f t "F 4 5 B t u / h r f t "F

35 Btu/hr 20 B t u / h r 23 B t u j h r 13 B t u / h r

0.56 x 10-6/"F 1.7 x 1OW6/"F

*ESTIMATED **GRAPHITE NOT WET BY FUEL SALT AT REACTOR CONDITIONS

FIGURF: 3.

141 l b s / f t 3 0.47 B t u / l b OF 0.83 B t u / h r f t "F 19 l b s / f t hr 10.7

HASTELLOY-N SPECIFIC GRAVITY THERMAL CONDUCTIVITY 'AT 1200 SPECIFIC HEAT AT 1200°F MEAN COEFFICIENT OF THERMAL EXPANSION (70-1 200°F)

DENSITY SPECIFIC HEAT (1200°F) THERMAL CONDUCTIVITY WITH GRAIN AT 68°F NORMAL TO GRAIN AT 68OF WITH GRAIN AT 1200°F NORMAL TO GRAIN AT 1200°F COEFFICIENT OF THERMAL EXPANSION WITH GRAIN AT 68°F NORMAL TO GRAIN AT 68OF

% % %

REACTOR PARAMETERS AND PHYSICAL PROPERTIES

ft ft ft ft

OF "F OF* OF*

FIGURE 4.

ONE-FIFTH SCALE CORE MODEL

hid

c i r c u l a t i n g f'uel.

l e variables of

e-model are then r e l a t e d t o

those of t h e reactor by t h e following proportionalities. ( l i n e a r dimensions IMSRE

= 5 ( l i n e a r dimensions)Model

( f l u i d velocity)Msm = ( f l u i d velocity)Model

(Reynolds Number)MSRE = (Reynolds Number)Model ( f l u i d age)MSRE=

5 ( f l u i d age)Model

( r e l a t i v e f l u i d pressure gradients i n ft-fluid),,

= ( r e l a t i v e f l u i d pressure gradients i n ft-fluid)Moael (turbulent heat t r a n s f e r

e f f i c i e n t s IMsRE

= 0.63 x (turbulent heat t r a n s f e r coefficients)Model Since t h e model was so small, not every surface and channel of t h e reactor core exposed t o salt w a s simulated.

As a r e s u l t , t h e model did not

@,ve exact comparisons, and w a s used only as a rough desi@ checking device t o e s t a b l i s h early i n t h e program the acceptability of major core concepts. For example, it was used f o r studies of t h e volute design and spacing of

f l a w distribution o r i f i c e s , design of s w i r l k i l l i n g vanes i n lower head, and preliminary measurements of solids s e t t l i n g characteristics of t h e lower head.

The full scale MSRE c

actual reactor.

Figure 5 is a photograph of t h i s model.

s t r u c t e d of carbon s t e e l

The model i s con-

of the moderator support

exception of t h e core blocks and p a r t h are of aluminum. The core blocks were

made by extruding alumin

mately t o shape including t h e 'longitudinal

grooves and then taking on

i s h cut on t h e 4 side surfaces.

t h e tolerances of t h e r e f o r economic reasons. In reduce t h e cost i f t h dynamics.

The vessel

t h e volute so t h a t t h r

1 i s almost an exact duplicate of t h e

Several transparent p l a s t i c w a l l cooling annulus and vo

Most of

increased (normally doubled) i n t h e model t i o n , other simplifications were made t o sumed t o have a s m a l l effect on t h e f l u i d ructed with a large g i r t h flange j u s t over could be removed with r e l a t i v e ease. ow6 were placed i n t h e vessel heads, core for viewing. Numerous holes were d r i l l e d

i n t o t h e vessel w a l l s a t various places f o r f l u i d measuring probes.

carbon s t e e l loop was b u i l t t o operate t h i s model and consisted of a pump,

FIGURE 5.

FULL SCALE CORE MODEL

gate valve t o control

flow, o r i f i c e flowmeter, loop cooler, 5400 gal.

surge tank and an ion

ange system f o r removing salt injected during

f l u i d age measurements.

All the i n i t i a l data from the full scale model was taken with t h e loop

r 1

f i l l e d with water and operated between

75OF and 80O~. This r e s u l t s i n

a Reynolds Number f o r t h e model about four times t h a t of t h e reactor.

Later

i n the program a thickening agent (Jaguar-508 by Stein, H a l l & Co.) w a s added t o t h e system i n order t o simulate Reynolds Nmibers, and t h e masurerents

which were strong functions of t h e Reynolds Number, were repeated.

Although Jaguar-508 imparts non-Newtonian c h a r a c t e r i s t i c t o the water, i n t h e low concentrations t h a t were used i n these t e s t s , t h i s was a negligible

As data i s presented i n t h i s report, it w i l l be noted whether o r not exact Reynolds Number s i m i l a r i t y existed. Items t o simulate consideration.

t h e t h r e e control rods and t h e surveillance specimen holder were not included i n the core model because t h e i r design w a s not s u f f i c i e n t l y well known when t h e model was b u i l t .

Rather, t h e regular graphite matrix was

continued throughout t h i s regions. DESCRIPTION AND ANALYSIS OF TEST RESULTS

Volute and Core Wall Cooling Annulus The main f u e l loop piping i n t h e MSRE i s 5 i n . Schedule 40. J u s t p r i o r t o entering t h e core vessel volute t h e pipe s i z e i s increased t o 6 i n .

The

cross sectional f l o w area of t h e 6 i n . pipe i s approximately equal t o that

of t h e volute, 28.8 in.* and 26.0 in.2, respectively.

The volute i s a con-

s t a n t flow area type and t h e t a i l end i s hydraulically connected t o t h e head end so t h e r e is recirculation.

One c h a r a c t e r i s t i c of t h i s type of volute

i s t h e variable s t a t i c pressure around it; therefore, i n order t o obtain

a uniform angular flow d i s t r i b u t i o n , it i s necessary t o use o r i f i c e s between the volute and core w a l l cooling annulus.

i n stacks of 3. I A

apart.

The o r i f i c e s a r e 3/4 i n . and occur

A t t h e head end of the volute t h e o r i f i c e stacks a r e 5 deg

A t t h e t a i l end of t h e volute , because of t h e lower f u e l velocity

and r e s u l t i n g higher s t a t i c pressure, t h e o r i f i c e stack spacing i s increased t o 22 1 / 2 deg. This o r i f i c e d i s t r i b u t i o n was determined from t h e 1/5 scale

10 model. The o r i f i c e holes are d r i l l e d at an angle of 30 deg with t h e tangent i n order t o maintain a tangential velocity component i n the annulus. The r e s u l t i n g high heat t r a n s f e r coefficients cool t h e core vessel w a l l and t h e reactor core can. Figure 6 i s a p l o t of t h e experimentally measured centerline v e l o c i t i e s i n t h e volute as a function of angular position and at various flow r a t e s . A t 1200 gpm water approaches t h e core through t h e 5 i n . pipe at a mean velocity of 19.2 ft/sec.

Immediately inside t h e volute t h e velocity jumps

t o about 23 f t / s e c because of recirculation around t h e volute.

t a i l end of the volute t h e velocity i s about 10 f%/sec.

A t the

The centerline

v e l o c i t i e s cannot be taken as absolute representations of flow rate, part i c u l a r l y at t h e i n l e t where two f l u i d streams of

mrge.

different velocities

Nevertheless, t h e l i n e a r decrease i n velocity i s a good indication

t h a t t h e water i s distributed uniformly t o t h e core w a l l cooling annulus.

A t the head end of t h e volute t h e man velocity through t h e o r i f i c e s i s

about 4 f t / s e c and at t h e t a i l end of t h e volute t h e mean velocity through t h e o r i f i c e s i s about 18 ft/sec.

Figure 7 i s a p l o t of t h e centerline

velocity i n the core w a l l cooling annulus as a function of elevation. Note t h a t t h e velocity decreases as t h e water moves down t h e annulus, Figure 8 i s a p l o t of t h e

because t h e t a n g e n t i a l component i s attenuated.

centerline velocity at t h e bottom of t h e core w a l l cooling annulus as a function of angular position around t h e core.

Note t h a t it i s quite f l a t ,

indicating uniform f l o w t o t h e reactor vessel lower head.

6, 7 and 8 w a s taken with water i n t h e loop.

Data f o r Figures

The Reynolds Numbers involved

i n t h e volute and core w a l l cooling annulus are so high (over 10 i n all cases at 1200 gpm) t h a t exact Reynolds s i m i l a r i t y i s not important, and t h e reactor vessel containing fuel salt w i l l have t h e same velocity profiles. A t t h i s point it would be informative t o compute t h e temperature dif-

ference between t h e bulk salt i n t h e core wall cooling annulus and t h e vessel

w a l l (Hastelluy N ) .

A t the midplane of t h e core (about 30 i n . up i n

the annulus) and at 1200 gpm, t h e f l u i d velocity i n t h e annulus i s 7.2 Ft/sec

(Figure 7).

Now, molten salt behaves as a conventional Newtonian f l u i d so t h a t

standard heat t r a n s f e r relationships may be used.

From t h e Dittus-Boelter

equation and with physical properties from Figure 3 we can compute a heat

ORNL-DWG 64-6724Af

20 u)

eJ16 w I-

I- 42 a

>. k

> 0

-I LA

0 6, ANGLE FROM INLET TANGENT (deg)

FIGURE 6.

ANGULAR DISTRIBUTION OF FLUID VELOCITY AT CENTER LINE OF VOLUTE

ORNL-DWG 64-6723A

70 ?

.-c

z z a a 50

z -I 8 0

-I J

W E

g I

0 J

8 i=

> 0

8 FLUID VELOCITY AT ANNULUS

(fps)

FIGURE 7. VERTICAL DISTRIBUTION O F F L U I D VELOCITY AT CENTER OF CORE WALL COOLING ANNULUS

ORNL-DWG70-11789 6

I 0

I 0 u

1200 gpm

I c)

600 gpm

135

180

~225

270

315

360

ANGLE AROUNDCORE (deg)

FIGURE8. ANGULAR DISTRIBUTIONOF FLUID VELOCITY AROUND BOTTOM OF COREWALLCOOLINGANNULUS

14 t r a n s f e r coefficient a t t h i s position i n t h e core wall cooling annulus

The heat generation rate i n the

(1i n . t h i c k ) of 1370 Btu/hr-ft2-OF.

3 vessel w a l l at t h i s point has been estimated t o be about 0.2 watts/cm The vessel w a l l i s 9/16 i n . t h i c k so t h e heat f l u x t o t h e s a l t , assuming t h e outside surface i s insulated, w i l l be 905 Btu/hr-ft

The temperature

drop across t h e f l u i d boundary l a y e r w i l l be only 0 . 6 6 ~ ~ N . ow the tempera-

ture drop i n t h e vessel w a l l with an i n t e r n a l heat source, and assuming again t h a t t h e outside surface i s insulated, i s represented by 5

2 k

AT = Q t' where:

t = w a l l thickness = 9/16 i n . Q = i n t e r n a l heat source = 0.2 w/cm 3

k = thermal conductivity

11.71 Btu/hr-ft-OF

Evaluating gives a temperature drop i n t h e metal w a l l of

1 . 8 1 O ~ .

Therefore

t h e o v e r a l l temperature drop from t h e outside surface of t h e vessel w a l l t o the salt i n t h e core w a l l cooling annulus i s t h e sum of t h e above o r only 2.47OF.

It i s beyond t h e scope of t h i s report t o include many d e t a i l e d

thermal analyses such as above.

These analyses have been made and many a r e

repo.rted i n References 8 and 9.

It was f e l t however t h a t one such computation

would be worthwhile t o give t h e reader an idea of t h e order of magnitude of these e f f e c t s .

Because of i t s r a t h e r low power density, l a t e r a l temperature

gradients are quite low in the MSFLE. Reactor Vessel Lower Head The lower plenum of t h e core vessel is f o m d by a standard 60 i n . OD ASME flanged and dished head, containing anti-swirl vanes and a drain l i n e

configuration.

The anti-swirl vanes consist of

48 p l a t e s s t a r t i n g about

2 i n . up i n t h e core w a l l cooling annulus and extending along radial l i n e s i n t o t h e lower head f o r about 38%of t h e radial distance t o t h e core centerline.

They are s l i g h t l y elevated off t h e core vessel w a l l , thus eliminating

as much corner area as possible where settled s o l i d s could accumulate.

The

vessel drain consist of a short section of 1 1 / 2 i n . pipe extending s l i g h t l y up i n t o t h e vessel head at t h e centerline, and having a conical umbrella over

it.

I n addition it incorporates a secondary drain which i s a 1/2 i n . con-

c e n t r i c tube coming up t h e middle of t h e primary drain, penetrating through

15 t h e drain pipe w a l l j u s t inside t h e vessel and wrapping around the drain pipe horizontally about 90".

The conical umbrella w i l l prevent gross

s e t t l i n g of s o l i d p a r t i c l e s (should they exist) i n t o t h e primary drain l i n e , however, Fuel containing solids may s t i l l drif't i n and out of t h e umbrella and over a long period of time, s o l i d s could s t i l l plug t h i s drain.

The

1 / 2 i n , l i n e serves as a s a f e t y drain since it i s designed t o prevent slow

migration of fuel i n and out.

Because of i t s s i z e , it would drain t h e

reactor t o o slow under normal conditions.

The secondary drain terminates

inside t h e primary drain j u s t below t h e freeze valve. The object of t h e anti-swirl vanes is t o prevent t h e s w i r l generated i n the volute from penetrating i n t o t h e lower head and creating an excessive radial pressure gradient.

The uniformity of f u e l flow through t h e graphite

moderator region i s a d i r e c t function of t h i s pressure gradlent.

Figure 9

i s a p l o t of t h e experimentally observed r a d i a l pressure gradient as measured by wall pressure taps.

Since the^ flow i n the lower head i s 3-

dimensional, t h e s t a t i c pressure at t h e wall i s not an absolute measure of the pressure influencing flow through t h e moderator region, nevertheless

it i s a good indication.

Note t h a t t h e pressure is s l i g h t l y higher at t h e

center, therefore, one would expect a s l i g h t l y higher flow through t h e modera t o r near t h e center.

This w a s measured t o be the case as w i l l be pointed

out i n t h e section on t h e moderator.

As t h e water goes through t h e anti-swirl vanes and heads toward t h e vessel centerline, it is turned by t h e lower head and produces a high velocity j e t adjacent t o t h e w a l l . Figures10 is a p r o f i l e of t h i s j e t measured at a radius of 17 inches and at 4 positions 90" apart. The flow r a t e w a s 1200 gpm.

Again, since t h e Reynolds Number i s so high, t h e same j e t w i l l

e x i s t with salt i n t h e vessel.

This j e t does not p e r s i s t much f a r t h e r towards

t h e vessel centerline when t h e f l a w characiter becomes gusty but s t i l l remains

turbulent.

From t h i s velocity p r o f i l e a heat t r a n s f e r coefficient can be

estimated by assuming p a r a l l e

l a t e geometry.

would be 4 times t h e distance

om t h e wall t o t h e peak velocity.

The equivalent diameter The cal-

culation with t h e Dittus-Boelter equation yields a heat t r a n s f e r coefficient 2 f o r a 1 salt of 540 Btu/hr-ft -OF. Heat t r a n s f e r coefficients were a l s o measured with a l o c a l l y developed "heat meter." on one surface

Basically, it i s an aluminum cube with a e l e c t r i c heater and a thermocouple i n i t s i n t e r i o r .

The surface opposite

i, ORNL-DWG 64-6722A1

ANNULUS

\PRESSURE

TAPS

0.4 0-

FLOW RATE = 1200 gpm

-0.2 -

FIGURE

9. RADIAL PRESSURE GRADIENT AT WALL OF LOWER VESSEL HEAD

ORNL-DWG 6 4 - 6 7 2 7 A 1

2.5

2 .o

1.5

DATA P( NTS MEASURED 90째 APART

FLOW RATE = 1200 gpm -

. 1.0

0 FLUID VELOCITY (ft/sec) FIGURE 10. VELOCITY P R O F I L E AT WALL J E T I N LOWER VESSEL HEAD AT RADIUS OF 17 IN.

la the heater i s then exposed t o t h e f l u i d stream and mounted flush with

t h e vessel surface. vessel w a l l s .

The heat meter i s thermally insulated from t h e

It is i l l u s t r a t e d and described i n more d e t a i l i n R e f . 10.

By measuring t h e e l e c t r i c a l power input and t h e temperature difference

between t h e meter and t h e f l u i d , t h e heat t r a n s f e r coefficient can be calculated.

The technique y i e l d s a heat t r a n s f e r coefficient i n a thermal

entrance region and it i s necessary t o convert them t o the "far downstream" This can be done with correlations i n References 4, 5 and

case.

addition, the measured coefficients must be converted from water t o salt with t h e Dittus-Boelter equation.

Heat t r a n s f e r coefficients were measured

at a radius of 17 i n . and 4 i n . with w a t e r i n the loop and were checked

with water thickened with Jaguar f o r Reynolds N-er equal.

s i m i l a r i t y and found

The data, after being converted t o t h e far downstream case and from

water t o s a l t , are shown i n Figure 11. Also shown on t h i s p l o t i s t h e heat t r a n s f e r coefficient as calculated f r o m the wall j e t . of t h e curve indicates turbulent f l a w .

Note t h a t t h e slope

More confidence i s placed i n t h e

data at 17 i n . than at 4 i n . because t h e flow i s well defined at t h e larger radius and less defined and gusty a t t h e s h o r t e r radius.

It i s ex-

pected t h a t the predicted values of these heat t r a n s f e r coefficients a r e conservative because they do not i n c l u e t h e e f f e c t of thermal convection which i s expected t o contribute roughly 100 Btu/hr-ft2-OF coefficient

t o t h e overall

Consideration w a s given t o t h e p o s s i b i l i t y of a source of nuclear power o s c i l l a t i o n s i n the MSRE core e x i s t i n g because of f u e l short c i r c u i t i n g i n t h e lower head.

If a f r a c t i o n of t h e f u e l d i d s h o r t c i r c u i t ,

then a corresponding amount of fuel must reside a l i t t l e longer i n the head.

This results i n t h e short c i r c u i t i n g fraction entering-the moderator

and l i t t l e cooler than t h e mean and t h e short-circuited fraction entering the moderator and a l i t t l e h o t t e r than t h e mean.

Power o s c i l l a t i o n s result

because of t h e f u e l temperature coefficient of r e a c t i v i t y .

Tests were

conducted on the model by i n j e c t i n g a conductive s a l i n e solution at a constant r a t e i n t o t h e lower head at various locations and measuring t h e o s c i l l a t i o n s i n e l e c t r i c a l conductivity at t h e vessel o u t l e t .

These measure-

ments indicate t h a t t h e power o s c i l l a t i o n s o r i g i n a t i n g from t h i s mechanism should be well below 0.01% of t h e mean power.

ORNL-DWG 70-1 1790 1000

DIUS OF 17 i n . ,

A CALCULATED FROM VELOCITY PROFILE OF WALL JET AT RADIUS OF 17 in.

LL 0

L r \ a c, m Y

500

I-

z w

2 LL LL

0 V

g 200 LL

II-

100

200

500

1000

2000

FLOW RATE (gpm)

FIGURE 11. HEAT TRANSFER COEFFICIENTS IN REACTOR VESSEL LOWER HEAD

Graphite Moderator Assembly

I n flowing through t h e reactor moderator assembly, f u e l first passes

through a moderator support structure and then through t h e moderator core blocks.

The support structure consists of two assemblies, t h e lower of

which i s a Hastelloy-N cross s t r u c t u r e (Figure 1) which i s t h e main support structure f o r t h e graphite.

Resting on t h i s i s a g r i d consisting of two

layers of rectangular graphite b a r s , one layer r e s t i n g on t h e other and perpendicular t o it.' "he purpose of t h e graphite a s s e d l y i s t o position and hold t h e graphite core blocks, and t o compensate f o r a difference i n thermal expansion between t h e Hastelloy N and graphite.

The r e s u l t i n g

square passages i n t h e graphite g r i d are small and t h e f u e l velocity i s h i @ (approx. 4 1/2 f%/sec), and therefore i t s pressure drop is high. Comparatively, t h e salt velocity through t h e bulk moderator f u e l channels i s l o w (approx. 0.7 f%/sec) and i t s pressure drop i s low. Figure 12 shows t h e experimentally determined head loss across t h e bodera t o r assenibly as measured i n t h e model.

A s m a l l correction has been applied

t o t h i s data t o account for small holes d r i l l e d through t h e support g r i d of t h e MSRE, as w i l l be discussed l a t e r i n t h i s section.

Also shown i n

t h i s figure i s t h e head l o s s across t h e support grid as measured i n a separate model made f o r t h i s purpose.

"he difference between these curves i s

t h e head l o s s i n t h e moderator channels themselves including entrance and e e t losses.

For these t e s t s , t h e loop w a s f i l l e d with water.

Note t h a t

t h e slope of t h e curves i n Figure 12 i s 2.0 as would be expected since t h e Reynolds N u n h e r with water i s over 4000 i n t h e f u e l channels and t h e bulk of t h e head loss i s due t o form losses i n t h e graphite grid.

With

f u e l s a l t t h e bulk of t h e head loss would s t i l l be from form losses.

The

Reynolds N W e r i n t h e f u e l channels would be approximately 1000 so t h e f l o w character would t h e o r e t i c a l l y be laminar, however, much of t h e tur-

bulence generated by f u e l passing through t h e tortuous i n l e t configuration would p e r s i s t through t h e f u e l channels.

-As a result one might expect a

very s l i g h t l y lower slope t o t h e curves i n Figure 12 with f u e l s a l t i n t h e -,

system.

As pointed out, t h e bulk of t h e pressure drop through t h e moderator assenibly i s due t o t h e gpaphike grid.

There i s a l s o very l i t t l e room f o r

cross f l o w i n the space between t h e graphite grid and t h e entrance t o t h e

ORNL-DWG 64-6724A1

1.0

0.5

0.2

0.1

0.05

0.02 100

200

500

1000

2000

5000

10,000

FLUID FLOW RATE ( g p m ) FIGUFE 12.

FUEL SALT PFZSSURE DROP ACROSS MODERATOR ASSEMBLY

fuel channels.

Therefore t h e uniformity of flow d i s t r i b u t i o n

cross

the moderator assembly is controlled t o a large extent by t h e uniformity of flow through t h e o r i f i c e s i n t h e graphite grid.

The f l o w d i s t r i b u t i o n

among t h e core fuel channels w a s experimentally measured by two different techniques.

The most successful technique was t o simply i n j e c t a small

amount of e l e c t r i c a l l y conductive salt solution i n t o t h e water flowing i n

a fuel channel and measure t h e time required t o pass two conductivity probes a given distance apart.

The device used consisted of a length of

1 / 4 i n . tubing blanked off on t h e end, and a small hole d r i l l e d i n t o t h e side where t h e s a l t solution w a s injected.

The tubing was i n s e r t e d down

i n t o a fuel channel about 6 i n . from the top.

Access t o t h e f u e l channels

w a s available through t h e viewing ports i n t h e upper head of t h e vessel.

The assenibly w a s made t o readily s l i p i n t o and out of t h e f u e l channels, and w a s s m a l l enough i n cross section so t h a t i t s presence would not sign i f i c a n t l y change t h e flow rate.

The conductivity probes were about 3 i n .

apart and about 2 i n . up from t h e injection hole. i n a s p e c i a l t e s t fixture.

The device w a s calibrated

The f l o w rate was mebsured i n 77 more o r less

randomly chosen passages and t h e data appear i n Figure 1 3 as a function of t h e core radius.

Note first t h a t 2 separate regions are present which represents

channels t h a t are mutually perpendicular t o each other.

These regions are

characterized by d i f f e r e n t i n l e t configurations because of t h e i r position over t h e graphite grid.

A plan v i e w of t h e graphite g r i d with t h e position

of t h e f u e l channels superimposed i s included i n t h e figure.

A separate

p l a s t i c model w a s b u i l t of a f e w 'adjacent f'uel channels with t h e graphite g r i d simulated t o investigate methods of correcting t h i s problem.

Based

on these t e s t s , it w a s determined t h a t 0.104 i n . holes (no. 37 d r i l l ) d r i l l e d through t h e upper graphite g r i d b a r and d i r e c t l y under the starved channel would e q u i l i b r a t e t h e flows. holes i s shown i n Figure 13.

A t y p i c a l location of one of these

Note a l s o from t h i s figure t h a t t h e flow r a t e

decreases s l i g h t l y with increasing radius.

This w a s expected and i s due

t o t h e radial pressure gradient i n t h e lower head as discussed e a r l i e r

(see Figure 9).

Actually t h i s flow d i s t r i b u t i o n i s b e n e f i c i a l because

t h e flaw rate i s highest where t h e power density i s highest, r e s u l t i n g M

i n a b e t t e r thermal u t i l i z a t i o n of t h e fuel.

In a large power producing

reactor, f o r instance, one would strive t o match t h e flow d i s t r i b u t i o n

ORNL-DWG 64-6719A!

2.8

' I

CHANNELS DESIGNATED BY 0

( 2.4

/CHANNELS DESIGNATED BY I

PLAN VIEW OF CORE SUPPORT BARS WITH FUEL CHANNELS SUPERIMPOSED

E a

- 2.0 0

-I

z z

U I 0

1.6

TYPICAL HOLE THROUGH SUPPORT STRUCTURE TO GET MORE FUEL TO THE STARVED CHANNEL

REACTOR CORE ICAN+

5 W

i.2

0 -I LL

0.8

// //

0.4

I DATA TAKEN BY DIFFERENT TECHNIQUE IN ANNULUS BETWEEN GRAPHITE AND MODERATOR SHELL I I I

RADIUS OF CHANNEL FROM CORE CENTER LINE (in.)

FIGURE 13.

RADIAL FLOW DISTRIBUTION OF FUEL SALT IN RTEL CHANNELS

iaJ

w i t h t h e power d i s t r i b u t i o n so t h a t t h e f u e l o u t l e t temperature across

Lastly note from Figure 13, t h e

t h e core would be.approximately equal.

rather large s c a t t e r of data points around t h e least square l i n e s .

This

i s a result of inherent inaccuracies i n t h e flow instrument and also because of tolerances i n t h e o r i f i c e s formed by t h e graphite grid.

Recall

t h a t the flow d i s t r i b u t i o n throu& t h e moderator i s controlled by these

o r i f i c e s because t h e i r pressure drop i s h i & and cross flow above them t o level out possible pertubations i s low.

The tolerances i n making t h e

model were normally double those of t h e reactor, so t h a t a greater variation i n o r i f i c e width w a s expected.

The nominal o r i f i c e width i s 0.375 in.

the model we measured these o r i f i c e w i d t h s with go nogo gages as high as 0.400 i n . and as l a w as 0.350 i n .

Estimates indicate t h a t t h i s variation

i s enough t o account f o r a l a r g e f r a c t i o n of t h e s c a t t e r i n Figure 13. It w a s therefore concluded t h a t t h e large s c a t t e r i n flow d i s t r i b u t i o n experienced i n t h e model w i l l not be present i n t h e reactor. The second technique used t o measure flow rates through t h e moderator channels was t o i n s t a l l pressure t a p s i n t h e aluminum core blocks and measure t h e pressure drop.

Pressure taps were i n s t a l l e d on 2 sides of 9

core blocks yielding information on 18 f u e l channels. calibrated before i n s t a l l a t i o n i n the model.

The core blocks were

Data from t h i s technique

w a s consistent w i t h date obtained from t h e salt i n j e c t i o n technique, although

with more s c a t t e r .

The data points at t h e extreme radius of t h e moderator

are shown as squares i n Figure 13. They represent the flow rate between t h e graphite moderator and reactor core can.

The u n i t s are given as gpm which

i s misleading, because t h e hydraulic parameters (e.g.,

area) i n t h i s flow region are not accurately known.

cross s e c t i o n a l f l o w To be s p e c i f i c t h i s i s

t h e flow rate corresponding t o t h e measured pressure drop i n a standard f u e l channel.

The s i g n i f i c a n t observation is t h a t these data points f a l l

more o r less on an extension o f t h e least squares l i n e . In t h e MSRE and i n t h e model, t h e regular p a t t e r n of t h e graphite support g r i d i s discontinued near t h e centerline of t h e core where t h e control rods and t h e surveillance s p e c i m n t holder are located.

greater salt v e l o c i t i e s past these components f o r cooling.

This allows

As noted

previously however, t h e control rods and specimen holder were not simulated i n t h e model.

Instead, t h e regular p a t t e r n of core blocks was continued

25 throughout t h i s region.

In t h e model, about 16 regulm fuel channels wereâ&#x20AC;&#x2122;

d i r e c t l y affected and a few more were i n d i r e c t l y affected by t h e disconThe average flow r a t e measured through these channels by the conductivity probe technique was 2.3 gpm. Two values measured i n tinued support grid.

t h i s region by t h e calibrated f i e 1 channel technique were 3.6 and 3.7 gpm. Fluid v e l o c i t i e s i n units of f t / s e c would be about t h e same i f control rods and t h e surveillance specimen holder were present. This i s more than adequate f o r cooling s o no attempt was made t o resolve t h e difference i n flow r a t e measured by t h e two different techniques. A suggested scheme f o r core power o s c i l l a t i o n s i s due t o changes i n the character of f l o w i n the centermost 16 f u e l channels. The Reynolds

N-er

of these channels i s i n t h e order of 3000 and i s therefore i n t h e

t r a n s i t i o n a l region.

It was speculated t h a t t h e character of flow could o s c i l l a t e from l d n a r t o turbulent between p a r a l l e l channels. To t e s t

t h i s hypothesis, a f l o w visualization study w a s made using an o p t i c a l l y

birefringent solution of Milling Y e l l o w dye i n water.

An almost full

s c a l e model of 4 p a r a l l e l fhel channels with i n l e t condit5ons similar t o t h e reactor was b u i l t from transparent p l a s t i c s . With two mutually perpendicular polarized p l a t e s , we were able t o observe clearly t h e t r a n s i t i o n between laminar and turbulent f l o w . The t r a n s i t i o n was smooth and continuous, and no o s c i l l a t o r y actson could be detected e i t h e r i n a single channel o r between channels t h a t could be regarded as significant. Many detailed temperature distribution calculations have been made f o r t h e MSRE core based on t h e r e s u l t s presented i n t h i s report. Results of these analyses a r e presented i n Refs. 8 and

9 and will not

be presented

here.

Reactor Vessel Upper Head lenum i s similar t o t h e lower plenum i n that it i s formed from a standard ASME flanged and dished head. The f u e l across t h e d i m t e r from t h e moderator f u e l enters t h e upper head unifo r a d i a l l y t o t h e center and leaves t h e core channels. The fuel then The reactor vessel upper

through a 10 i n . o u t l e t p i vortex i n t o t h e o u t l e t pipe.

The model showed no tendency of t h e water t o

A f t e r t h e model w a s b u i l t , a s t r a i n e r s t r u c t u r e

(Figure 1) was added t o the reactor o u t l e t design which penetrated down i n t o t h e upper plenum. Its purpose was t o catch graphite chips (graphite f l o a t s i n f u e l s a l t ) should they break loose from t h e moderator. This

26 device was not i n s t a l l e d i n t h e model but it would not be expected t o influence t h e flow significantly. Velocity p r o f i l e s i n t h e model w e r e determined i n t h e upper head by a f l u i d age technique.

E l e c t r i c a l l y conductive salt' solutions were injected

as a s t e p function at t h e core i n l e t .

Conductivity probes were placed at

selected locations i n t h e upper head presumably along t h e f l u i d streamlines. By measuring the time required f o r these conductive pulses t o pass between

t h e probes, t h e average velocity between t h e probes can be calculated. There are, of course, inaccuracies i n t h e calculation because they assume knowledge of t h e streamlines

Velocity p r o f i l e s estimated by t h i s technique appear i n

streamlines. Figure 14.

and they neglect l a t e r a l mixing between

Plotted i s t h e average velocity between the probes indicated

(see sketch) as a function of distance from t h e vessel w a l l , assuming t h a t t h e streamlines

were p a r a l l e l t o the w a l l .

A t t h e rated flow r a t e

(1200 g p m ) , t h e Reynolds Nuniber i n %he upper head is high for both w a t e r

and f u e l salt and t h e f l o w i s turbulent.

Heat t r a n s f e r coefficients

t o t h e vessel w a l l s can therefore be estimated w i t h Dittus-Boelter type equations.

Assuming p a r a l l e l p l a t e geometry where t h e equivalent diameter

i s twice t h e channel width, t h e estimated heat t r a n s f e r coefficient i s a

l i t t l e over 100 Btu/hr-ft2-OF neglecting thermal convection. This value has been generally confirmed with heat meter measurements taken at approximately t h e same location.

The calculation i s assumed conservative because

t h e e n t i r e upper head would be a thermal entrance region ( l o w length/diameter

ratio).

The region near the knee of t h e upper head w i l l have a lower heat

t r a n s f e r coefficient but t h e heat generation i n t h e MSRE vessel w a l l i n t h i s region is very low.

I n addition, some s m a l l channels have been milled

i n t h e MSRE core support flange t o allow a small flow of cool f u e l t o short c i r c u i t d i r e c t l y f r o m t h e core w a l l cooling annulus t o t h i s region o f t h e upper head.

The flow rate i n t h i s c i r c u i t has been estimated t o be between

25 and 50 gpm depending on certain manufacturing 'and assembly tolerances. Estimated temperatures i n t h i s region are presented i n Reference 9 .

Miscellaneous Measurements The o v e r a l l head l o s s through t h e core vessel f r o m t h e 5 i n . i n l e t l i n e t o t h e 5 in. o u t l e t l i n e w a s measured and i s shown i n Figure 15.

The

curve has a slope of 2.0 indicating that t h e head l o s s is primarily due t o form drag as expected, and very l i t t l e due t o skin f r i c t i o n .

. ORNL-DWG 70-11791 c

/OUTLET

PIPE

PROBE LOCATIONS

0.1

0.2

MODERATOR CORE BLOCKS

0.3

0.4

0.5

0.6

0.7

MEAN VELOCITY (ft/sec)

2.5 h

2.0

'0.79

Y .r

-t

1.5

0.1

0.2

0.3

0.4

0.5

0.6

0.7

MEAN VELOCITY (ft/sec)

FIGURE 14.

FLUID VELOCITY PROFILES I N REACTOR VESSEL UPPER HEAD

ORNL DWG 64-672OAl

(00

200

500

1000 2000 FLOW RATE (gpm)

5000

OVERALL FLUID HEAD LOSS ACROSS MSRE CORE FROM 5 INCH INLET PIPE TO 5 INCH OUTLET PIPE

FIGURE 15.

~0,000

Experiments were conducted t o obtain information on t h e behavior of heavy s o l i d p a r t i c l e s i n t h e f u e l s a l t .

The question arose because

if s u f f i c i e n t oxygen o r water vapor come i n contact w i t h f u e l s a l t , z i r -

conium oxide would p r e c i p i t a t e .

A considerable e f f o r t was made i n t h e MSRE

t o prevent t h i s , and indeed it never did happen.

Nevertheless, during

t h e design stage it was desirable t o know t h e disposiâ&#x201A;Źion of heavy s o l i d p a r t i c l e s should they occur.

One of t h e most l i k e l y places f o r p a r t i c l e s

t o s e t t l e out i s i n the lower head of t h e reactor.

This i s also one of

t h e more c r i t i c a l areas because they could p o t e n t i a l l y develop i n t o a s i g n i f i c a n t heat source on t h e primary containment surface.

There was

a l s o concern t h a t i f enough p a r t i c l e s developed, they may plug up t h e drain l i n e configuration. The experiments consisted of adding 2 o r 3 pounds of sized iron

f i l i n g s t o t h e water a t t h e core i n l e t while the pump was running. They were added slowly over a specified period of time, generally about 20 minutes. The pump was then turned off e i t h e r immediately after t h e addition o r l e f t running f o r an additional period up t o 8 hours. After t h e pump w a s turned o f f , t h e loop w a s drained and t h e quantity of s o l i d s t h a t remained i n t h e reactor vessel lower head was determined.

The s o l i d s passed

through t h e core only once because downstream of the core model was t h e 5400 gallon surge tank i n which t h e p a r t i c l e s would most c e r t a i n l y s e t t l e

out.

All data were taken a t 1200 gpm, mostly with water but some runs

with thickening agent .added for Reynolds Number simulation.

A summary

of t h e data is shown i n Figure 16. A p a r t i c l e diameter of 200-300 microns seems t o be c r i t i c a l , t h a t i s , p a r t i c l e s l a r g e r than t h i s s e t t l e d out i n significant q u a n t i t i e s , and p a r t i c l e s smaller than t h i s passed through t h e system.

The gusty nature of t h e f l o w around t h e drain l i n e assembly a l s o

seems t o have some a b i l i t y t o re-suspend p a r t i c l e s l e s s than 300 microns i n s i z e if t h e pump i s kept running a f t e r t h e addition phase.

The s o l i d s

t h a t did deposit i n t h e lower head had t h e appearance of an annular r i n g

around t h e drain pipe but d i d not touch it.

In no case w a s significant

quantities of s o l i d s found i n t h e 1 1 / 2 i n . drain pipe.

Because of t h e

way t h e model w a s b u i l t , it w a s not possible t o check t h e 1/2 in. emergency i

drain tube after each run.

After a l l runs were completed, however, t h i s

tube was examined and no p a r t i c l e s were found i n it.

After all t h e runs

NOMINAL MESH LOOP S I Z E OF SCREEN LIQUID IRON APERTURE VISCOSITY FILINGS. .jmicrons) ( l b / f t hr) 100

40 20 150 150

150 50-30 50-30 50-1 00 50-1 00

I I

147 370 '833 104 104 248 248 104 290 4 90 290-490 290-1 47 290-147

2.0 2.0 2.0 2.0 2.0

9 9 9 9

2.0 2.0 2.0 1/2* 1/2* 1/2* 1/2*

% OF ADDED SOLIDS THAT WERE RECOVERED WT SOLIDS FROM VESSEL ADDED LOWER LEAD Iser?l(Ib) (%)

TIME INTERNAL PUMP RUN USED TIME AFTER LOOP TO ADD SOLIDS WERE FLOW SOL IDS ADDED RATE

(min)

(hr)

120 120 120 35 32 20 20 20 20 20 20 20

0 0 0 0

1180 1180 1180 1180

3 3 3 3

1180

1180 1180 1180 1180 1180 1180 1180

3 3 3

61%

8 0 0

8 0

2 2 2 2

%O%

45% 55% 9%

55% 5.6% 56% 44% 30% 0%

THICKENING AGENT ADDED TO WATER FOR REYNOLDS NUMBER SIMULATION

FIGURE 16.

SUMMARY OF SOLIDS ADDITION DATA

31 were completed, t h e top head was removed from the vessel and it was found t h a t a large amount of f i l i n g s had s e t t l e d f a i r l y uniformly on They had rusted together, otherwise, a

t h e t o p of t h e core blocks.

large fraction of them may have f a l l e n back down i n t o t h e f u e l channels when t h e pump was stopped.

Note t h a t t h e top of t h e core blocks i s i n

the shape of a pyramid (Figure 2 ) . Calculations have been made f o r t h e MSRE t o determine t h e e f f e c t of precipitated fuel p a r t i c l e s on t h e con-

tainment w a l l temperature near the drain pipe.

They a r e not serious and

some of t h e r e s u l t s are reported i n Reference 9. Experiments were performed t o determine t h e behavior of gas bubbles i n t h e core. inlet.

Bubbles between 1/8 and 1/4 i n . were i n j e c t e d i n t o t h e core

A t 1200 gpm v i r t u a l l y all t h e bubbles went dam t h e core w a l l

cooling annulus and e s s e n t i a l l y none short c i r c u i t e d through t h e slots i n t h e core support flange t o t h e upper head.

In contrast t o t h i s , a t about

350 gpm, all t h e bubbles went through t h e core support flange s l o t s and none went down t h e core wall cooling annulus.

The bubbles injected at t h i s

luwer flow r a t e were considerably larger, however, than a t t h e higher flow rate.

A t 1200 gpm, no pockets could be noted anywhere i n the core vessel

where bubbles tended t o accumulate.

In t h e MSRE, bubbles i n t h e f u e l loop

t h a t have been generated by t h e xenon s t r i p p e r spray r i n g are estimated t o be i n t h e order of 0.010 in.

Certainly they w i l l t r a v e l with t h e s a l t

at 1200 gpm and not accumulate i n pockets o r short c i r c u i t through t h e

core support flange. EXPERIENCE WITH THE MSRE

A b r i e f summary of t h e MSRE operating h i s t o r y i s as follows:

First criticality

- June

1, 1965

Nuclear operation terminated Time reactor w a s c r i t i c a l

1 2 , 1969

- 17,655 hours

- 13,172 equivalent full power hours f u e l loop - 21,788 hours.

Nuclear heat production Salt circulating i n

- Dec.

32 There were no d i r e c t f l u i d measuring probes of any kind (velocity, pressure drop, e t c . ) i n s t a l l e d i n t h e MSRE core. Even t h e flow rate was

an i n f e r r e d quantity from measured AT'S i n t h e f u e l and coolant loop and t h e coolant s a l t flow r a t e from a venturi.

The only instrumentation t h a t could

y i e l d f l u i d dynamic performance information i n t h e core was 64 thermocouples spotted at various locations on t h e outside of t h e core vessel (not i n wells).

These thermocouples yielded no indications of any adverse conditions

inside t h e core.

Another r a t h e r i n d i r e c t source of information would be by

t h e r e a c t i v i t y balance, which was kept i n d e t a i l f o r t h i s reactor because of i t s experimental nature.

Again, no effects were noted t h a t could be

a t t r i b u t e d t o fluid dynamics i n the core.

The conclusion then i s t h a t ,

based on lack of evidence t o t h e contrary, t h e MSRE core d i d behave as w a s expected f'rom t e s t i n g t h e model.

It should be noted however, t h a t

because o f t h e conservative design of t h e core, a r a t h e r large malfunction would have been necessary i n order t o see it w i t h available instrumentation. REFERENCES

Robertson, R. C., MSRE G s i g n and Operations Report, Part I , Descript i o n of t h e Reactor Design, USAEC Report ORNL-TM-728, January 1965.

Entire Issue of Nuclear Applications and Technology, Vol. 8, No. 2 , February 1970.

Molten-Salt Reactor Program Semiannual Progress Report f o r Period Ending July 31, 1964, USAEC Report ORNL-3708, November 1964.

Aladyev, I. T., Experimental Determination of Local and Mean Coefficients of Heat Transfer f o r Turbulent Flow i n Pipes, Tech. Memo 1356, NACA.

Sparrow, E. M. , e t al. , Turbulent Heat Transfer i n a Thermal Entrance Region of a Pipe with Uniform Heat Flux, Appl. Sci. Res., Sec. A, VOl.

6 . Abbrecht, P. H., The Thermal Entrance Region i n N l y Developed Turbulent Flow, AIChE Journal, Vol. 6 , No. 2 , June 1960. 7.

Lindauer, R. B. , Revisions t o MSRE Design Data Sheets , Issue No. 9 , ORNL Publication ORNL-CF-64-6-4 3, June 24, 1964, I n t e r n a l Distribution Only. Engel, J. R. and Haubenrefch, P. N . , Temperatures i n t h e MSRE Core During Steady-State Power Operation , USAEC Report ORNL-TM-378 , November 5 , 1962.

9. Bettis , E. S : , e t ak., WRE Component Desipp Report, ORNL Report MSR-61-67, June 20, 1961, I n t e r n a l Dfstribution Only. 10. Mott, J. E., Hydrodynamic and Heat Transfer Tests of a N l Scale R e - E n t r a n t Core Aw. 8 , 1958, __- - ~ ORNL Publication ORNL-CF-58-8-54 I n t e r n a l Distribubbon Only. I

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