ORNL-2095.txt

   

 

ORNL-2095

. C-85 :‘Reactors—.Aircmft Nucliear Propulsion Systems

This decument consists of 198 pages.

Cop;&g?#‘of 300 copies. Series A,

AIRCRAFT REACTOR ENGINEERING DIVISION

DESIGN REPORT ON THE AIRCRAFT REACTOR TEST

A. P. Fraas and A. W, Savolainen

May 1956

DATE ISSUED

 

 

OAK RIDGE NATIONAL LABORATORY
Operated by
UNION CARBIDE NUCLEAR COMP ARY
A Division of Union Corbide and Carben Corporation
Post Office Box X
Oak Ridge, Tennessee

RESTRICTED DATA

This decument contains Ré'st'f‘i'iyz;‘éefi“‘-putq as defined in the Atomic
Energy Act of 1954, [ts tronsniittal or the disclosure of its contents
s . . - ; ol s
in any manner to an unauthgrized person is prohibited.
» ’ RO

. .

  
W@ NN W

. A, Abattieilo
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. B. Bachulis
. J. Barton

. E. Bealle

. Bender

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. L. Bech

. 4. Borkowski
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. E. Boyd

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. B. Briggs

. E. Brewning
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. B. Callihan
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. 5. Carlsmith

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E. Center (K-25)

. A, Charpie
. B. Clarke
. E. Clifford
. G, Cobb

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. B. Cottretl
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C-85 — Reactors—Aircroft Nuclear Propulsion Sysiems

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OMe-T<TMEBPAINZLEIITFIMZ

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. 94. J. P. Murray (Y-12) 124. W. J. Stelzman
95. M. L. Nelsen 125. E. Storte
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. 116. C. Sisman 146. C. D. Zerby
117, J. Sites 147-156. ORNL - Y-12 Technical Library
118. M. J. Skinner Document Reference Section
119. G. M. Slaughter 157-189. Laboratory Records Department
120. G. P. Smith 190. Laboratory Records, ORNL R, C,
121. P. G. Smith 191-193, Central Research Library
122. A. H. Sneli

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g 220. Burecu of Aercnaoutics

 
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¥as TN P
ER sy, B O
T e _
TS ey gy TV AN A wp
cemea e oL g e

e e e

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. 'SIEBR.ET"

FOREWORD

The Aircraft Reactor Test {(ART) program was launched early in 1954, The reasoning
and experiments that led first to the choice of the circulating-flucride-fuel reactor concept
and then to the reflector-moderated reoctor configuration have been presented in previous
documents. These include a presentation by R. C. Briant to the USAF Advisory Com-
mittee, outlining the cbjective and status of the ORNL-ANP Program (ORNL report
CF-53-2-126), summaries of the preliminary design work on the Aircraft Reactor Experi-
ment (ARE) (ORNL-1234) ond the operation of the ARE (ORNL-1845), and o compre-
hensive summary of ORNL-ANP aircraft power plant designs up to May 1954 (ORNL.-1727.
The preliminary layout of the ART and facility was given in the ART hazards report
(ORNL-1875) end in the ANP Project Quarterly Progress Report for the Period Ending
December 10, 1954 (ORNL-1816).

The information compiled in this design report is intended to present a fairly detailed
picture of the ART design as of its approaching completion. It is expected that the
design as defined in this report will be changed somewhat as information is derived

from component test experience, further analyticel work, or fabricational problems.

ACKNOWLEDGMENTS

Most of the material presented here has appeared previously in quarterly progress
reporfs, special ARE and ART reports, or ART design memorandums and data sheets,
The work of all members of the ANP staff is therefore represented here, and their contri-

butions are thus acknowledged.

 
 

 
SECRET

CONTENTS

 

PART |
D ESIGN R T E R A L ettt a s e e et e seare et eeaeabenes

Military Requitements ... ettt et ettt et e ebe st s et er e ere s
Reactor Design Criteria ..ottt

The Circulating-Fuel Reflector-Moderated Reactor ...,

PART il

THE AR T ASSE MBI Y e e
The Reactor Assembly .ottt et et
RO D IMEI STONS oo e et ettt e e ettt et e et e et et e et e eaee e et e eeneenne s
R @CTOr-MOdEIGIOr .. oo iveeiiiii e e ettt e e e et e e e r e aeaaan
Fuel Sy stem oo e b et e ek e et et et eae e eae e taeneeas

Core Hydrodynamics ..ottt e
Pumps and Expansion Tank ..ot
Fill-and-Drain System {Including Enricher) ...,
APl INg Sy S oM et e sttt saa
Fuel Recovery SysSTem ...ttt et ettt aeanea

Heat Exchanger and Heat DUmPS ..o
O oG0S SYSTOM ..ottt bttt te e ettt ae et e e et et e e nse e nneeaeas
Aircraft-Type Shield ...ttt eee s
Controls and INSHUmMenTAHION ..o ettt et r et aes
AUKTITGEY SYSTOIMS oo ettt ettt ettt et ettt ean e e

He UM SUDPlY e e ettt ettt ettt
NIrogen SUpPly SysStem .o ettt
Electrical Power System, Distribution, and Auxiliary Equipment ...
Fuel and Sodium Pump Lubricating and Cooling Oil Systems ...,
Process Water SyStem ..ottt e b et et

The FaCi Iy oottt ettt ettt et et e et e et

TRE BUildiNg oottt a ettt e
The Reactor Assembly Cell e
The Shielding Experiment Facility ..o e

DESIGN AND DEVELOPMENT STUDIES oot
Fuel Development ... ettt e e e e e ae et
Structural Material ..ottt
Corrosion of Structural Materials ...,
Radiation Effects on Structural Materials
Structural Design Analyses ..o,

ko The Reactor ..o o s
Auxiliary Components ...l o g i g bbb bbb

 

B W W
    

e,
Eee g
&y, i tosg, F e . o F
= i = B .
, e e, .. =,
e “ B B
L L E L X

N g, oy

Yot R e

Radiation Heating on the ART Equatorial Plane in the Vicinity of the
Fuel-to-NaK Heat EXChanger ..........coiiiice ettt e s

Radiation Heating in Various Regions of the North Head ...
Beta- and Gamma-Ray Activity in the Fuel-Expansion Chamber and the Cff-Gas System ...

PART IV

ENGINEERING TEST UNIT i et et ettt asr e et
Specific Test ObjECHIVES .
Warmup and Shakedown Testing oottt e
OPErOEING T@SES oottt ettt ettt et e ee et h et e eh e e s en bR e
Reactor ASSembly ..o e s

Reactor Disassembly ...t e

CONS T RUC T ION AND O ER A TION oot e e e ettt an e
Plans for Installation of the AR T o et e r e e e et e ettt et ee s taeaeeeaenenns
Operation of the ART .o e e e

Filling and Feating ..ottt et e et et e
Enriching 16 Critical oottt
Low-Power-LLevel Experiments . ...t
Operation At Power ...t e e eans

Plans for Disassembly ...

STATUS OF DESIGN AND DEVELOPMENT e

FLOW DEAGRAMS et e s e

APPENDIX B

PHOTOGRAPHS OF REACTOR MODELS AND STAGES IN THE CONSTRUCTION OF THE
AR T F A LT Y oo ettt et b e eb e

B B O G R AP H Y oo it bbb et

 

viil
Figure No,

D e W R e

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1

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LIST OF ILLUSTRATIONS

Title

Vertical Section Through Reactor Assembly

Horizontal Section Through Fuel Pump-—Expansion Tank Region
Horizontal Section Through Sodium Pumps

Vertical Section Through Lead-Water Shield

Effect of Reactor Dimensions on Concentration of U233 in Fuel

Effect of Reacter Dimensions and External Fuel Volume on Total U233
investment

Effect of Reactor Dimensions on Outside Pedk-to-Average Power-Density
Ratio in Core of Reflector-Moderated Reactor

Effect of Reactor Dimensions on Percentage of Fissions Coused by
Thermal Neutrons

Partictly Assembled Island Showing the Lower Half of the Inconel Core
Shell and the Upper Half of the Beryliium Reflector for the High-
Temperature Critical Assembly

QOuter Core Sheil and Partially Assembled Beryllium Reflector of the
High-Temperature Critical Assembly

Completed High-Temperature Critical Assembly of the Reflector-Moder-
ated Circulating-Fuel Reactor

Core Layout

Diagram of Core Flew System Showing Relations Between the Inlet
Headers, Inlet Guide Vanes, and Core

Section Through Fuel Pump—Expansion Tank Region Showing Xenon-
Removal System of the Fuel Pump

Schematic Diagram of Fuel Fill-and-Drain System, Including Enricher
Fuel Fill-and-Drain Tank Cooling System

Model of Main Fuel-to-NaK Heat Exchanger Channel

Prototype ART NaK-te-Air Radiator

Plan of the ART Building

Section of the ART Building

Reactor Assembly Celi

ART Shielding Experiment Facility

Pressure Load Distributions at Full Power

Conditions for Thermal Cycling Tests of One-Fourth-Scale Model of
ART Core Shell in Comparison with ART Thermal-Cycling Conditiens

ART Support Structure

Movement of Reactor Due to Expansion of NaK Lines

Layout of NaK Piping in Reactor Cell

NaK-to-Air Radiator, NoK Pump, and NaK Piping External to Cell
Fuel Fill-and-Drain

  
 

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2]

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& & & &

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Figure No.
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Flow Diagram

O~ O tn AW R e

Ny — DD

Figure No.

B.1-B.6
B.7-8.33

 

Title

Gamma-Ray Heating in the Vicinity of the Fuel-to-NaK Heat Exchanger
on the Equatorial Plane of the ART

Heating in Copper-Boron Leyer by Alpha Particles from the B'%(n,a)Li’
Reaction

Configuration of ART North Head Showing Members Referred to in
Table 9

Totsl Power and Power Density in the Gas Space of the ART as o
Function of the Gas Volume and the Helium Flow Rate for a Fuel
Flow Rote of 22 gpm

Power Density in the Off-Gas Line as @ Function of Time and Gas
Velume in the Expansion Tank for a Fuel Flow Rate of 22 gpm

North-Head Weldability Model Showing L.ower Deck and Peripheral Ring,
Step 1

North-Head Weldability Model, Step 2
North-Head Weldability Medel, Step 3
Nerth-Head Weldobility Model, Step 4
North-Head Weldability Model, Step 5
Nerth-Head Weldability Mode!, Step 6
North-Head Weldability Model, Showing Another View of Step 6

APPENDIX A

Fuel Fiit-and-Drain, Enriching, Sampling, and Recovery Systems
Sodium System

Cff-Gas System

Cell Pumps Hydraulic Drive Systems
Reactor Pumps Lube Oil System

Main, Auxiliory, and Special NaK Systems
NaK Pumps Lube Oil System

Process Air System

Helium System

Nitrogen System

Compressed Air System

Process Water System

AFPENDIX B
Title

Photographs of Reactor Models

 

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145171
Part |

DESIGN CRITERIA
 

 
DESIGN CRITERIA

MILITARY REQUIREMENTS

Studies made by Air Force contractors have indi-
cated that aircraft for missions involving strategic
bombing should be capable of operation at sea level
and o speed of approximately Mach 0.9, or at
55,000 ft (or higher) ot Mach 2.0 (or higher), or
above 85,000 ft at about Mach 0.9. An airplene of
unlimited range that could fly ony one or, even
better, twe or more of the possible sfrategic
missions would be extremely valuable if it became
available in the 1965 to 1970 era. In addition to
the strategic bomber spplication there are reguire-
ments for lower speed (Mach 0.4 to 0.9), manned,
nuclear-powered airplanes, such as radar picket
ships and patrol bombers, The problems associated
with supplying @ beachhead @ substantial distance
from the nearest advanced base indicate that a
logistics-carrier airplane of unlimited ronge would
alsc be of considerable value, The strategic-
bombing missions for manned aircraft with shielded
reactors have been deemed to be of such crucial
importance as to more than justify the development
cost of the nuclear power plant.

A nuclear power plant of sufficiently high per-
formance tc provide the power required for the most
siringent operating conditions would be able to
take care of any of the other manned-airplane
requirements. Design studies have indicated that
nuclear power plants capable of preducing from
100 to 300 Mw will be required. The 60-Mw Aircraft
Reactor Test (ART) is a logical and expeditious
intermediate step in the production of the required
high-power reactors and should give a reactor that
will be capable of providing sufficient power fo
operate a radaer picket ship, patrol bomber, or
logistics carrier. A reactor power of 60 Mw was
selected because it is approximafely the power
that must be reached for an investigation to be
made of the engineering problems that must be
selved and for disclosure of the operating charac-
teristics to be expected of the higher powered
reactors required for high-cltitude supersonic
strategic bombers. The size and weight of the
reactor and shield will conform with aircraft re-
quirements, and, inscfar as was possible in the
fimited time available, the design of the important
components has been based on concepts satis-
factory for airborne application.

REACTOR DESIGHN CRITERIA

The general requirements feliowed by the ORNL-
ANP project personnel have been consistert with
those for aircraft reactors set forth by the Technical
Advisory Board.! The Board stated that the re-
actor must have a power output of several hundred
megawatts, that it must heat air to temperatures
in excess of 1100°F, ond that the reactcr core
must occupy a space not exceeding a few feet in
finear dimensions. The choice of materials must
be limited, of course, to those with desirable
nuclear properties. It was taken into consideration
that the reactor structure must be able to with-
stand considerable accelerations or shocks without
large misalignments or changes in reactivity re-
sulting. A reactor with a mechanically simple,
rugged core would thus present a considerable
advantage. The only compensating factor in the
stringent requirements was considered to be the
short lifetime over which the reactor must operafe —
500 to 1000 hr.

The uranium investment per recctor was expected
to be a limiting factor in the size of the air fieet,
and therefore the uranium content was to be as
iow as possible. Shortening the reprocessing times
was expected fo be of major value in lessening
the uranium inventory required to maintain an air
fleet. Overriding the xenon effect after @ prolonged
shutdown or a radical reduction in power level was
expected to be difficult for reactors with high
specific power., However, for some reactors, such
as the homogeneous reactor, the xencn would be
removed shortly after formation and thus would
present no problem.

One major problem was considered to be the con-
struction of control mechanisms (and, to o (esser
degree, sensing instruments) for reliable operation
that would respond rapidly and not require undue
extension of the shielded volume. It was believed
that a liquid-fuel reactor with a strong negotive
temperature coefficient of reactfivity would be
self-stabilizing and would thus require a minimum
of control.

With regard to the vulnerability and scfety of the
nuclear-powered airplane it was recommended that
the reactor be isolated from engine failure by an

 

‘Rep-ort of the Technical Advisory Board to the
Technical Committee of the Aircrafi Nuclear Propulsion
Program, ANP-52 {Aug. 4, 1950).
intermediate tluid system., For a high-temperature,
high-power-density reactor in an airplane, loss of
the coolant or loss of circulstion of the coolant
would cause rapid overhecfing. Thus a system in
which the hect of the primary cooclant was frans-
mitted to the engines threugh an intermediate heat
exchanger would be advantageous.  Also, by
locating the intermedicte heat exchanger in a pro-
tected position, probably in the shield, it would be
possible to have short, protected flow lines for
the primary fluide The secondary lines to the
engines could then be independent of each other.
Special provisions would have to be made to take
care of afterheat from fission preducts after shut-
down.

THE CIRCULATING-FUEL REFLECTOR-
MODERATED REACTOR

The circulcting-fuel reflector-mederated reactor
was designed to meet the criteric described. The
use of circulating fused fluoride saits for the fuel
carrier and heat transfer medium has the important
advantcge of eliminating the heat fransfer stage
required in solid-fuel-element recctors within the
reactor core ond with it a complex, delicate matrix
of hect transfer surfaces. With the fuel as the
ptimary heat transfer medium, it has been possible
toc wrap the intermediate heat exchanger around the
spherical reflector-moderator and thus to keep the
heat exchonger within a relatively small shielded
volume. The octivation of the secondary coolant,
NaK, poses problems in ground hendling and mainte-
nance that are offset by the size and weight ad-
vantages of the circulating-fuel design.

The primary control mechanism of this type of
reactor is the streng, negetive temperature coeffi-
cient. Since it will be possible, by scrubbing the
fuel with helium in the pump expansion tank, to
remove the xenon that is produced, it will not be
necessary fo make provision for overriding the
xenon after a shutdown.

Although the circulating-fuel reactor has ¢ some-
what lorger uranium investment in the operating
reactor than some proposed aircraft reactors have,
if atlowances are made for the inventory in spore
fuel elements and in the reprocessing plont, the
tofal investment per airplane can be lower for the
circulating-fuel recctor than for most other types of
regetors. in addition, it will be much easier to
repleace the liguid fuel than to chonge the selid
fuel elements. Especially shielded or remotely
controlled ground-handling facilities will be re-
quired only for the fuel filling and draining opera-

tions. The reprocessing of the liquid fuel will be
much simpler and faster than the reprocessing and
refabrication of solid fuel elements, and the
shorter reprocessing time will, of course, lower
the uranium investment.

Operation of the Aircraft Reactor Experiment
(ARE)2 demonstrated that a high-temperature
circulating-fuel reactor could be built and operated
and that the materials and machinery which had
heen developed for cperation at elevated tempera-
tures were satisfactory, It showed that the pre-
dicted latge negative temperature coefficient of
reactivity and the resultant self-regulofory charae-
teristics of the reactor could be achieved. The
ARE and other successful experiments have thus
indicoted the high probability that aircraft nuclear
power plants employing circulating-fuel reactors
con be developed for militery application,

A careful analysis of the multitubular core-
moderated reactor configuration typified by the
ARE disclosed a number of sericus problems. The
high power densities necessary in a full-scale
aircraft reactor (1 to 5 kw per cubic centimeter of
core} give severe gumme and neutron heating con-
ditions in the moderating material and hence large
thermal stresses. [t was felt that cracking and
spalling of ¢ brittle material such as the beryliium
oxide used as the ARE moderator could be expected
and that designing to ollow for this would be
exceedingly awkward, The weight and drag penal-
ties associated with the use of a low-temperature
moderator such as water would be excessive, Ex-
tensive corrosion tests with hydroxides at reason-
able temperatures, that is, 1200°F, had shown that
none of the hydroxides would be satisfactory. Too
little was known about the hydrides for them to
be used as the basis for a design. A reacter con-
structed of graphite was toe large, and hence the
shield was toc heavy., The only moderatoreregion
materials combination that seemed to be compatible
with the fluoride fuel—Inconel system was sodium-
cooled beryllium.

The various detailed designs that were evolved
in the effort to incorporate these materials in @
multitubular core fall into twe groups. The first
group of designs was based on the use of a heavy
header sheet across the outiet face of the core to
carry the fluid pressure locds associated with both
the sodium and the fuel pressure drops. A light
header sheet was to be used across the inlet face

 

W, B. Cottrell et al., Operation of the Aircraft Reac-
tor Experiment, ORNL-1845 (Aug. 22, 1955).

 
 

that would be sufficiently flexible to accommodate
differential thermel expansion between the porallel
fuel tubes and between the core inlet and outlet
passage shrouds. The beryllium would be in the
form of hexagonal blocks, as in the ARE, and would
be spaced relative to the tubes by spiral wire
spacers fo ensure the proper sodium-flow-possage
thickness. The principal disadvantages of this
design approach were the severe thermal and pres-
sure stresses in the cutlet header sheet {which
would be at o high temperature} and the severe
pressure stresses in the inlet header sheet under
off-design conditions (e.g., one pump out} if the
sheet were made thin enough to accommodate the
requisite amount of differential thermal expansion.
Possibly less serious disadvantages included a
tendency of the fuel tubes to be bent by the beryl-
lium under the lateral accelerations to be expected
in flight.

fn an effort to evelve an arrangement with greater
promise, o second series of designs was prepared
that was based on large disks of beryllium being
placed normal to the tube axes. These disks were
designed to carry the loads induced by the fuel
and the sodium pressure drops. To ensure the
proper sodium flow distribution, the spacing be-
tween the fuel tubes and the beryilium was ogoin
accomplished by spiral wire spacers. Moderately
thick header sheets were employed, with provision
for both transverse and axial differenticl thermal
expansion at the outlet end. This latter feature
presented o major problem, and, to dote, nc realiy
sound detail design has been worked out to
accommodate both the differentiol expansion be-
tween the beryllium and the header sheet and at
the some time take the pressure loads imposed by
off-design conditions.

Multigroup calculations made concurrently with
the above design studies indicated that it was
advantageous to lump the fuel heavily to minimize
the amount of structural material in the reactor.
Concurrent shielding studies showed that about
12 in. of good neutron shielding fellewed by @
layer of boron was required between the core and
the hecat exchanger te keep the activation of a
fluid such as NaK from becoming excessive.
Pursuance of these design precepts yielded a
reflector-moderated configuration which gave not
only a lower shield weight than had been obtained
with any configuration previously devised but also

a lower fuel concentration than could be obtained
from a multitubular core having the same core
diameter, according te multigroup calculations
made ot the time, Even more important, these
calculations indicated thet the reflector-moderated
core design gave both the lowest average and the
fowest peck power density in the fuel for a given
core diameter of any core design studied.

The volume heat source in the circulating fuel
presents a set of problems having no pravious
technological parcilel, The fuel temperature in
stagnant regions tends to rise rapidly, and it can
easily reach excessive values if the hydredynamic
design is not satisfactory. Anaclytical sclutions
for a few simple geometries, such as fully de-
veloped flow in a straight tube, have been worked
out, and tests have validated the analytical soiu-
tions. More complex geometries, such as those
represented by the reflector-moderated reactor,
can be analyzed with assurance only by means of
experimental techniques. Thus a multitubular
reactor core design offers o major advantage in
that it is more amenable to analysis, although it
would be likely to present header flow possage
design problems ot the inlet and outlet,

A careful appraisal of the many different factors
invelved led finally to the choice of the reflector-
moderated type of core because it appeared to
give the lightest shield for ¢ given power density
and fuel concentratien in the fuel, as well as a
configuration relatively free of stress concentrae
tions in the core structure. Although the actual
fabrication ond cssembly of the concentric con-
tiguration of the core of the reflector-moderated
reactor present some exceedingly difficult prob-
lems, it is felt that they are no more difficult than
the problems that would be encountered in o well-
designed multitubular-core reacter. Thus, after
reviewing all the varicus considerations, it was
decided in 1954 that the reflector-moderated core
should be adopted as the basis for the ART design.

The objectives of building and operating the ART
are the investigation of the methods of construc-
tion end the estimated performance characteristics
of such a reflectoremoderated circulating-fuel
reactor. An operating life of 500 hr ot or near
60 Mw is considered to be a desiroble goal, The
information gained from the test will provide a
sound basis for the design of full-scale aircraft
reactors,
 

 
Part Il

THE ART ASSEMBLY

..............
 

 
e

THE ART ASSEMBLY

THE REACTGCR ASSEMBLY

The reactor ond shield assembly will consist
of the core, the beryllium island, the beryllium
reflector-moderator, the pressure shell, the fuel
system (including the fuel pumps ond the xencn-
removal system), the fuel-to-NaK heat exchanger,
the reflector-moderator cooling system, the control
system, and the aircraft-type shield. The reactor
(Fig. 1) comprises a series of conceniric shells,
ecch of which is a surface of revolution about the
vertical axis. The two inner shells surround the
fuel region at the center (that is, the core of the

~reactor) ond separate it from the beryllium islond

and the outer beryliium reflecter. The fuel circu-
lates downward through the bulbous region where
the fissioning tokes place and then downward and
outward to the entrance of the spherical-shell fuel-
to-NaK heat exchanger that lies between the re-
flector outer shell and the main pressure shell.
The fuel flows upward between the tubes in the
heat exchanger into two cenirifugal sump-type fuel
pumps at the top. From the pumps it is discharged
inward to the top of the annular passage leading to

the reactor core. The fuel pumps are located in

the expansion tank region at the top of the reactor,
A horizontal section through this region is shown
in Fig, 2.

The reflector-moderator is cooled by sodium
which flows downward through spaces between the
core shells and the beryllium and through passages
in the beryliium and back upward through the
annular spaces between the beryllium and the en-

. closing shells and between the pressure shell and
' the pressure shell liner. Two sump-type centrifugal

pumps at the top of the reactor circulate the sodium
first through the reflector-moderator and the island
and then through the small toroidal sodium-to-NoK

~ heat exchangers around the periphery of the pump-
“expansion tank region, A horizontal section through

this region is shown in Fig. 3.

The Inconel pressure shell constitutes both the
main strucfure of the reactor and a compact con-
tainer for the fuel circuit. The pressure shell
encloses a liner composed of ancther Inconel shell,
a layer of boron carbide to remove thermal neutrons,
and an inner lnconel can. Heat generated in the
pressure shell by the absorption of gamma rays
from the fuel will be remeved by sodium flowing
upward from the bottom of the island between the
outer surface of the liner and the inner surface of

the pressure sheil. Passages at the top of the
reactor will direct the sodium frém the outer
surface of the pressure shell to the sodium-to-NaK
heet exchanger inlets. The pressure shell is
separated by thermal insulation from the gamme-ray
shield, which is o layer of lead. The lead, in turn,
is surrounded by a region of borated wafer. A
vertical section through the lead-water shield is
shown in Fig. 4.

Data that give materials and operating charac-
teristics of the ART are presented in Table 1,
except for the system flow ond pump data, which
are given in Table 2, Key data on dimensions are
given in Table 3.

REACTOR DIMENSIONS

The dimensions of the reactor core were estabe
lished by calculations and by critical experiments.
A parametric study was made of a set of 48 related
reactors in which the parameters of core diameter,
reflector thickness, and fuel thickness were varied
over a wide range. ! In this study the effects of the
geometry of the reflector-moderated reactor on the
physical quantities of interest, such as eritical
mass, required mole per cent of uranium in the fuel,
and power distribution, were determined. Core radii
of 20, 30, 40, and 60 cm and extrapolated reflector
thicknesses of 20, 30, and 40 cm were selected for
the set of related reactors. Poison (sedium and
Inconel) distributions in the reflector and island
were obtained from a study by Bussard and others.?
The fluoride fuel NaF-ZrF‘éoUFA‘ was used for the
entire set of calculations. From the set of 48
reactors determined by the specified independent
parameters, it was possible to select 30 reactors
that would give results that could be used to cross-
plot and interpolate the properties of the 18 omitted
reactors fo an accuracy sufficient for this survey.
The results obtained are summarized in Figs. 5, 6,
7, and 8,

Physicol property studies indicated the desir-
abitity of limiting the uranium content of the fuel
te 5 mole % UF , (see Part llI, “'Design and Devel-
opment Studies®’), and therefore it is clear from

 

e, s. Burtnette, M. E. LaVerne, and C. B. Mills,
Reflector-Moderated=-Reactor Design Parameter Siudy,
Part I: Effect of Reactor Proportions, ORNL CF-54-7-5
(Nov. 8, 1954).

2R, W. Bussard et al., The Modefator Cooling System

écg z};e Reflector-Mcderated Reacior, ORNL-1517 (Jan, 22,
54},
 

ORNL-LR-OWG 1804

- FUEL PUMP

 

 

 
          
 

  

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= | &
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=2

 

 

 

 

     

“FILLER PLATES

      

  

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R3L .

ONTROL.

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oY _
SRz Fal
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e ul . el w
35 v Ul TR 1
[ o Z i \ C4 ; a
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uw = &y u I_ =
T c & Ul &=
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h i e < et
=
e

 

ical Section Through Reactor Assembly.

Vert

Fig. 1.

10
 

TABLE 1. ART DESIGN DATA

Power

Design heat output (kw)

Core heat flux

Core power density (max/avq)

Power density, maximum (kw per liter of core)

Specific power (kw per kg of fissicnable material in core)
Power generated in reflector (kw)

Power generated in island (kw)

Power generated in pressure shell (kw)

Power generated in lead layer (kw)

Power generated in water layer (kw)
Materials

Fuel

Reactor structure
Moderator
Reflector

Shield

Primory coclant

Reflector coolant

© Secondary eoclant

Fuel System Properties

Uranium enrichment (% UZBS)
Critical mass {kg of U23%)
Total uranium inventory (kg of U235)
Consumption at maximum power (g/day)
Design lifetime (hr)

Design time at maximum power (hr)

- Burnup in 500 hr at maximum power (%)

Fuel volume in core (Ffa)
Total fuel volume (Fts)

Meutron Flux Density in Core

2-=:lsen:"'v')

10% ev <E <107 ev (neutrons.cm™
Thermal < E < 164 ev (neufronsocm“2-sec-])
Thermal, maximum (neutrons-cm"z-sec"t)

2

Thermal, average (neutronsccm™ osec"i)

Control

Shim coniral

Rate of withdrawl

Temperature coefficient {cver-all)
Temperature coefficient (fast)
Thermal fissions (%)

Neutren leakage (%)

Prompt neutren [ifetime {usec)
k’eff (clean, as loaded)

Ak (temperature)

Ak (poisons)

Rggs {hot and peisoned)

Conversion ratie

60,00C

Heat transperted out by circulating fuel
2:3

1300

940

2040

600

210

132

4

E~4!<:E*:=-ZrF4-=UF4 (50-46-4 mole %)

Inconel
Beryllium
Beryflium

Lead and borated water
The circulating fuel
Sedium

Nak

93.4
23
64
88
1500
500
2.9
3.2
10

3% 1019
1x 1013
2% 1014
5x 1013

One rod of 5% AL/E
3.3 % 10~4% Ak/kesec
~2.3 % 16™% (Ak/k)/°F
~5 % 10~9 (Ar/R)/°F
40

32

400

1.04

0.004

0.036

1,00

11
TABLE 1 {continued)

 

Circuloting-Fuel=Coolant Systems

Fuel in core
Qutlet temperature {°F)
Temperature rise (°F)
Mean flow velocity (fps}

Reynolds number (for mean axial velocity)

Fuel-to-NaK heat exchanger
Volume of fuel (fl'a)
Yolume of NaK in tubes (f’ra)
VYolume of inconel in tubes (Hg)

Inconel tube surface in contact with fuel (fi‘z)

Heat exchanger thickness (in,)

Maximum temperature (°F)

Temperature drop (or rise) (CF)
Pressure drop {psi}

Flow rate (f?g/sec)

VYelocity through the tube mateix (fps)
Reynolds number

Heat tronsfer coefficient (Bfu/hr-hzofiF)

Cooling system for NaK-fuel coolant
Maximum air temperature (OF)
Ambient airflow through NeK radictors {cfm)
Radiator cir pressure drep (in. H20)
Blower power required (total for four blowers) (hp)

Tetal radiater inlet face area (hz)

Cocling system for moderator
Maximum temperature of sodium (°F)
Sodium temperature drop in heat exchanger (°F)
NoK temperature rise in heat exchanger (°F)
Pressure drop of sodium in heat exchanger {psi)
Pressure drop of NaK in heot exchanger {psi)
Flow rate of sodium through reflector (f?s/sec)
Flow rete of sodium through island and pressure shetl {fia/sec)
Flow velocity of sodium through reflector and island (fps)

Reynclds number of sodium in reflector and island

1600
350

5
85,000

2,45
2,97
1.88
903

3.25

Fuel NakK Coolant

1600 1500

350 430
3¢ 13

2,96 10.45

8.77 24

3740 120,000
2215 10,000

750
243,000
9

600

100

1250
200

250

7

7

1.35
0.53

30
170,860

 

TABLE 2. SYSTEM FLOW AND PUMP DATA

 

Fue! Sodium

NaK

{(Reflector Coolont} (Fuel Coclant} (Sodium Coolant)

NaK
{Fill-and- Drain
Tank Caoclant)

 

Number of pumps 2 2 4 2

Pumping head, ft 2737 100 215-360 270 360
Fiow per pump, gpm 645 440 1200 430

Pump speed, rpm 24062600 3200 26903400 30253400
Pump power per pump, hp 2227 61-118 38-48

 

12
TABLE 3. REACTOR DIMENSIONS

 

REACTOR CROSS-SECTION EQUATORIAL RADII {in.)

Control rod thimble

Inside 0.750

Thickness 0.062

Qutside 0.812
Sodium passage

Inside 0y

Thickness 0.130

Outside 0.942
Beryllium

Inside 0.942

Thickness 4.121

Qutside 5.063
Sodium passage at land

Inside 5.063

Thickness 0.188

QOutside 5.251
Inconel shell

Inside 5.251

Thickness 0.125
- Qutside 5,376
Fuel

Inside 5.376

Thickness 5.124

Qutside 10.5C0
Outer core shell

Inside 10.500

Thickness 0.125

Qutside 10.625

- Sedium passage at land

Inside 10.625

Thickness 0.188

Outside ' 10.813
Beryllium refiector

inside 10.813

Thickness 10,855

Qutside 21.648
Sodium passage

Inside 21.668

Thickness 0.125

Qutside 21.793
Inconel shell

Inside 21.793

Thickness 0.240

Qutside 22.033

 

Stainless-steel-clad copper——B4C cermet
inside
Stainless steel thickness
Copper- B4C thickness
Stainiess steel thickness

Qutside

Stainless-steel-canned B4C
Can
Inside
Thickness
Qutside
B4C tile
Inside
Thickness
Qutside
Shim gap
Can
Inside
Thickness
Outside

Shim gap

Outer reflector shell
Inside
Thickness
Cutside (max)

Channel

Tangent to first tube

Tube radius

Center line, first tube
Twelve spaces at 0.250
Center line, thirteenth tube
Tube radius

Circle tangent to thirteenth tube
Spacer

Gap

Channel

Inside
Thickness
Cutside

Gop
Inside
Thickness
Outside

22.033
0.C10
0.080
0.010

22,133

22,733
0.005
22,738

22,738
0.240
22,378
0.029

22.407
0.005
22.412
c.o21

22,433
0.262

22.495

22.500
22,510
0.115
22.625
3.000
25.625
0.115
25,740
0.008
6.022

25.770
0.120
25.890

25.890 -

0.030
25.920

13
TABLE 3

REACTOR DIMENSIONS

 

14

Boron jacket
Inside
Thickness
Outside

B4C tile
Inside
Thickness
Ouytside

Pressure shell liner
Inside
Thickness
Outside

Sedium gap
Inside
Thickness
Dutside

Pressure shell
Inside
Thickness
Outside
CORE

Diameter (inside of outer shell at

equator}, in.
island outside diameter, in.
Core infet cutside diameter, in.
Core inlet inside diameter, in.
Core inlet areaq, in.2

Core equatorial cross-sectional areq,

2

in,

25.920
0.062
25.982

25.982
0.328
26.310

26.310
0.375
26,685

26.685
G.125
26.810

26,810
1.000
27.810

21

10.75
il
6.81
58.7
256.2

REFLECTOR-MODERATOR REGICN

Volume of beryilium plus fuel, ft3
Yolume of beryliium only, £
Cooling passage diemeter, in.
Number of passages in island

Number of passages in reflector

FUEL SYSTEM

Fuel volume, f3
In 36-in.-long core
fri inlet end outlet ducts
In expansion tank when }2 in. deep

in heat exchanger

28.2
24.99
C.187
120
288

3.21
1.410
C.08
2.84

 

 

7 pump volutes

Total in main circuit

Fuel expansion tank
Volume {8%), fe3
Widsh, in.

Length, in.

SODIUM SYSTEM

Sadium velume, 13
{n expansien tank
ln annular passage at pressure shell
In reflector passages (total)
in first deck
It pump and heat exchanger
In second deck
In island passages (fotal}

Tetalin main circuit

Inside diameter of sodium transfer tube

to reflecter, in.

Inside dieameter of sodium transfer tube

from reflector, in.

Inside diameter of sodium transfer tube

to istand, in.

Area of sodium passage to reflector,
2

ine

Area of sedium passage from reflector,

2

in.

Areo of sodium passage to island, in, 2

FUEL-TO-NaK HEAT EXCHANGER

Tube center-line—to—center-line

spacing, in.

Tube ocutside diameter, im

Tube inside diameter, in.

Tube wall thickness, in.

Tube spacer thickness, in

Mean tube length, in.

Equatorial crossing angle

Inlet and outlet pipe inside diameter,
in.

Inlet and outlet pipe outside diameter,
in.

Nurmber of tube bundles

Number of tubes per bundle, 13 x 20

Total number of tubes

0.84
8.38

0.5787
13.625
32.500

0.16
1,60
0.90
0.47
0.35
0.42
0.44
4.34
2,375

3.875

1.437

4.426

5.847

1.619

0.250

0.229 o 0.231

0.180
0.025
0.020
65.000
26°20
2.469

2.875

12
260
3120 E
TABLE 3 (continued)

 

Center-line radius of NaK inlet pipes

Center-line radius of NaK outlet pipes

19.590
19.590

SODIUM-TO-NaK HEAT EXCHANGER

Tube center-line~to—center-line

spacing, in.

Tube outside diameter, in.

Tube inside diometer, in.

Tube wall thickness, in.

Tube spacer thickness, in.

Mean tube length, in.

Number of bundles

Number of tubes per bundle, 15 % 20

Total number of tubes

Inlet and outlet pipe inside diameter,
in.

Inlet and outlet pipe outside dicmeter,

in.

PUMP ~EXPANSION TANK REGION

Vertical distance above equator, in.
Floor of fuel pump inlet passage
Bottom of lower deck
Top of lower deck
Bottom of upper deck
Center line of fue! pump discharge
Center line of sodium pump discharge
Top inside of fuel expansion tank
inside of dome
Qutside of dome
Top inside of sodium expansion tank
Top outside of sodium expansicn tank
Top of fue!l pump mounting flange

Top of sedium pump mounting flange

Dome radius, in.
inside
QOutside

FUEL PUMPS

Center-line—to—~center=line

spacing, in.
VYolute chamber height, in.
Estimated impeller wéigh’r, b
Critical speed, rpm
. Shaft diameter, in.

Shaft everhang, in.

0.2175

0.1875
0.1375
0.025
0.030
28

2

300
600
2.469

2.875

17.625
19.125
19.656
24.000
21.437
26.125
29,25

29.875
30.875
34.312
34.812
47.000
50.243

29.875
30.875

 

 

 

Distance between bearings, in.
Impeller diameter, in.

Impeiler discharge height, in,
Impeller inlet diameter, in.
Shaft length {over-all), in.

Shaft cutside diameter between

bearings, in.

Lower bearing journal outside

diameter, in.

Shaft outside diameter below seal, in.

Thrust bearing height from equator, in.
Number of vanes in impeller

Diameter of top positicning ring, in.
Diameter of bottom pesitioning ring, in.

Quter diameter of top flange, in.

SODIUM PUMP

Center=line—~to«center-line

spacing, in.
Volute chamber height, in.
Estimated impelier weight, ib
Criticel speed, rpm
Shaft diameter, in.

Center-line lower bearing to center-line

impeller, in.
Distance between bearings, in.
Impeller diameter, in.
Impeller discharge height, in,
impeller inlet diameter (ID), in.

Shaft length (over-all), in.

Shaft outside diameter between

bearings, in.

L ower bearing journal cutside

diameter, in.

Shaft outside diameter below seal, in.

Thrust bearing height above equatsr, in,

Number of impeller vanes
Diometer of top positioning ring, in
Diameter of bottom positioning ring, in.

Qutside diameter of top fiange, in.

23.000

2.500
10
6000+
2.250
13.3

12
3.750
0.250
3.500
31.5
2,375

3.400

2.25
51.907
10
6.200
6.190
10.000

 

15
FUEL TO
CORE

&
Vo
1
Y

3
Vo

7
-f:.;%

 

CRNL-LR-DWG 8835

X

SO

R
AR
R

NN S‘

 

N
R
SN,
T

R
N

T T T R T
AT TR

N
AN

N
3,

Fig. 2. Horizontal Section Through Fuel Pump-Expansien Tank Region.

Fig. 5 that considerable advantage is gained by
increasing the reflector thickness from 20 te 30 cm
but little odvantage from o further increase to
40 cm. Even more important, the activotion of the
Nal in the heat exchanger can be reduced markedly
by increasing the reflector thickness. A 30-cm-
thick reflector gave a good compromise between
shielding, criticality, and NaK activation consider-
ations., The remaining analyses of the data were
therefore restricted to a 30-cm reflector thickness.,

The effect on total uranium investment of the
reactor dimensions and the external fuel volume,
such as will exist in the heat exchanger external

16

to the core, is shown in Fig. 6. The surfoce shown
for zero external volume is, of course, simply that
for critical mass. At @ core radius of 30 cm the
critical mass is virtually independent of fuel thick-
ness, In the surfaces for both the 2- and 4-ft3
external volumes, simple visuel examination indi-

cates the probable existence of an optimum set of

proportions. The reversal of scales for the 8-ft3

surface waes necessitated by the existence of an
extreme peck at what was the front corner.

The effect of reactor dimensions on the peak-to-
average power-density ratic is shown in Fig. 7.
The peak power density for the reflector-moderated

5
 

—
—

_\_\

S
Y

  
 

 

SR
ST
S
<%
o

 
 
 

TN ‘\'?‘\“
T aly

 

NoFLOW TO ™.
MODERATOR AND ISLAND -~

    

.~ CONTROL ROD

 
  
 

ORNL-_R-DWG 8837

NN
Na RETURN FROM \ \\
MODERATOR AN

  
 
  
       
  

  
  
  
 

 

   

THIMBLE

 

A‘;% //
S ;s /
3 @’} . / //

T T T ST N ,
T e NS
. Y
. 5 /
/‘ S L ,;'; )//
P N A/
i 7w N s
s [ Nak W ray
A WINLET 0
) - \\ // ’/
- S ,-/
,
ay
y
~ 2
- P
~. ,/
. e /
. .. .
S ,/
=
PR
T
o

Fig. 3. Horizontal Section Through Sodium Pumps.

reactor occurs af the outside shell of the fuel
annulus., The effects of reactor dimensions on the
ratic are small, as can be seen; an increase in core
radius from 20 to 60 cm at a constant fuel thickness
gives a decrease in the ratio of about 20%, while
an increase in the fyel thickness from 5 to 20 ¢m
-at a constant core radius results in an increase in
" the ratio of about 33%.

The percentage of fissions caused by thermal
neutrons is shown as a function of reactor dimen-
sions in Fig. 8. The least-thermal reactor (28%)

: has the thickest fuel layer and the smallest core,

 

ok

while the most-thermal reactor (45%) has the thin-
nest fuel layer and the largest core. The reactor
core dimensions established on the basis of this
study are presented in Table 3.

A series of room-temperature critical exper ments
was set up for checking the caleulations.® These
reflector-moderated assemblies were of simple ge-

ometry, and materials variations were made to

 

3D. Scott and B. L. Greenstreet, Reflector-Moderated
Critical Assembly Experimental Program, ORNL CF-54-
4-53 {April 8, 1954),

17
 

Lo =
ORNL-LH-2WG 16457 .
- GONTROL RO
-

   
   
   
 
 
  
 
  
 
  
  
 

 

_~—ALUMINUM TANK

7

P 18-in.-THIGK BORATED WATER SHIELD
P

FUEL PUMP DRIVE HYDRAULIC MOTORS

P HEXAGONAL CANS FILLED WITH LiCH
REACTOR SUPPORT PADS

-~ PARAFFIN
ALUMINUM TANK—

~10~in-THICK BORATED
SUPPORT FOR WATER SHIELD

WATER SHIELD
HANGER RODS

SUPPORTING LEAD SHIELD

 

I

       
  
  
  
  

STEEL SUPFORT STRUGTURE FOR
THE REACTOR ASSEMEBLY AND THE

LEAD AND WATER SHIELDS
4.3-in—THIGK LEAD

GAMMA-RAY SHIELD -

_— ALUMINUM TANK
Yy=in.- THICK
AIR GAP -+

32-in-THICK BCRATED
WATER SHIELD

Vi - THIGK /

-- COCLING WATER FASSAGES,
PHERMAL INSULATON —

34 -in~0D TUBES ON
4-in. CENTERS

HEXAGONAL CANS FILILED WITE LICH et / 3-in~-THICK LEAD GAMMA-RAY SHiELD
Yomin - THICK THERMAL INSULATION -~/

— FUEL DRAIN LINE

Fig. 4. Vertical Section Through Lead-Water Shield.

 

18
7

REFLECTOR THICKNESS = 20 CM

>
<2

77

7

UEIE CONMOCENTRATION IMOLE %)
o o

 

SEERET
ORNL-LR-DWG 1177

 

REFLECTOR THICKNESS =30 CM

 

REFLECTOR THICKNESS =40 CM

Fig. 5. Effect of Reactor Dimensions on Concentration of

check consistency with theory and the funda-
menfal constants, The first assembly was a basic
reflector-moderated reactor with two regions — fuel
and reflector. The fuel region contained uranium
and a fluorocarbon plastic, Teflon, to simulate the
fluoride fuels, and the reflector region contained
beryllium. The fuel region was constructed in the
shape of a rhombocuboctahedron (essentially, a
cube with the edges and corners cut away to give

U235 in Fuel.

cctagenal cross sections) to approximate a sphere
within the limitations imposed by the shape of the
available beryllium. The system was made critical
with 24.35 [b of U235,

The assembly was then modified to include three
regions with a beryllium island separated from the
reflector by the fuel. The first of the three.region
assemblies had no Inconel core shells, the second
included Inconel core shells without end duels, and

19
EXTRAPOLATED REFLECTOR
THICKNESS = 30 CM

FOTAL YEID fWESTMENT (LB}

 

EXTERNAL FUEL VOLUME = 4 FT°

EEERET
ORNL~LR-DWG 1176

  

 

 

EXTERNAL FUEL VOLUME = 8 FT3

Fig. 6. Effect of Reactor Dimensions and External Fuel Volume on Total U235 fpvestment.

the final one mocked up the reactor, including the
end ducts. The results of these experiments are
presented in Table 4 and in reports on the indi-
vidual experiments (see "‘Bibliography’’).

A tow-nuclear-power, high-tempercture critical ex-
petiment was also performed.? The reactor section

 

4A. D. Calithan et al., ANP Quar. Prog. Rep. Sept.
10, 1955, ORNL-1947, p 58.

20

of the assembly closely resembled the current de-
sign of the ART in that it included the annular fuel
region separated from the beryllium island and
reflector by Efg-in.ufhick inconel core shells of the
proper shape. Photographs of the assembly are
shown in Figs. 9, 10, and 11. The mockup differed
from the ART principally in that the fuel was not
circulated and there was no sedium in the reflector-
moderator regions. The system was filled initially,
 

SHECRET
! HNL-t R DWG
EXTRAPGLATFD REFLECTCR CHNL DWG 1196
THICKNESS — 20 ¢m

  
  

POWESRR DENSITY RATIO

y
/ T B 7

 

0
o
§;§ —
T
K
0
D
g p
o N e
s . . - -~
g o o
’74‘;&,/7 B e o5 -
e e
>y iy - < 23
e, e @
e s TS
< g |

Fig. 7. Effect of Reactor Dimensions on Outside
Peak-to-Average Power-Density Ratic in Core of
Reflector-Moderated Reactor.

SRoRET
AR ] R~ WG 47
EXTRAPOLATED REFLEGTOR ORNL-[.R=-DWG 't

THICKNESS = 30 cm

[eed

<2

CER CENT THERMAL FISSIONS

 

<& - "~ N 3 - )
2y s .
iz, -
s o .
fs;":;f;p & P 5>
e L

Fig. 8. Effect of Reacter Dimensions on Per-
centage of Fissions Caused by Thermal Neutrons.

for cleaning ond testing, with a 50-50 mole %
mixture of molten NaF and ZrF,. Increments of
molten Na,UF, (with the uranium enriched to 93%
U233) were then added to the NaF-ZrF , mixture in
the sump tank. After each addition of Na UF ,, the
mixfure was pressurized into the core and then

draired. This procedure was continued until the

critical concentration was attained,
The control safety rod was located within a
1.50-in.-ID Inconel thimble along the vertical axis

 

Fig. 9. Particlly Assembled lsiand Showing the
Lower Half of the inconel Core Shell and the Upper
Half of the Beryllium Reflector for the High-
Temperature Critical Assembly.

of the beryllium island. The rod was a cylirdrical
annulus of a neutron-absorber compact with @ den-
sity of 6.5 g/cm3, The principal coastituents of
the compact were Sm,0, (63.8 wi %) and (5d, 0,
(26.3 wt %); the outside diameter of the absorber
section was 1.28 in., and the gnnulus was % in.
wide. The system operated isothermally at 1200°F,
normally, but with the electrical heaters available

the temperature could be raised to 1350°F.

The critical fuel concentration was found to be
6.30 wi % {2.87 mole %) uranium, and the excess
reactivity was about 0.13% Ak/k. The over-ali
temperature coefficient of reactivity between 1150
and 1350°F was shown to be negative and tc have
a value of 2 x 105 (AL/k)/°F. An increase in the
uranium concentration of the fuel from 6.30 to
6.88 wt % resulted in an increase in reactivity of
1.3% Ak/k. The control rod had a value of 1.7%
Ak/k when inserted tc o point 4 in. above the

2}
TABLE 4. COMPOSITIONS AND DIMENSIONS OF REFLECTOR-MODERATED-REACTOR CRITICAL ASSEMBLIES

 

Three-Region Assembly with Core Shells of

 

Three-Region Assembly with }’S-in.-Thick

 

k|6,'°ifi"’7"'|’ii‘:k }fls'i""ThECk }é‘i"'"ThiCk inconel Core Shells end End Ducts
Aluminum laconel Inconel
Assembly number CA-200 CA-20b CA-2c CA-211 CA-21-2 CA-22 CA-23
Beryllium island
Volume, 2 0.37 0.37 0.37 1.27 1.27 1.27 1.75
Avercge radius, in.
Spherizal section 5.18 5.18 5.18 5.18 5.18 5.18 7.19
End ducts 3.86 3.86 3.86 3.86
Mass, kg 19.4 19.4 19.4 67.0 67.0 67.9 92.2
Fuel region (excluding shells and
interface plates)
Volume, £ 1.78 178 1.72 2.06 2.06 247 1.56
liters 50.4 50,4 48.8 58.3 58.3 76.0 44,3
Average radius, in.
Sphericol section
Inside £.24 5.24 5.31 5.31 5.31 5.31 7.32
Outside 2.51 9.51 2.44 9.44 9.44 9.44 9.44
End ducis
inside 3.99 3.99 3.99 3.99
Outside 5.28 5.28 679  5.28
Distance between fuel sheets, in. 0.639 0.284 0,142 0.142 0.173 0.142 a.142
Mess of components
Teflon, kg 99.38 99.27 94,37 108.88 108.18 126,77 81.54
Urarium loading, kg 5.00 11.74 22.07 26.02 21.57 30.45 19.97
U235 oading, kg £.66 16.94 20.56 24,24 20.67 28.35  18.62
Uranium densify,b g/cn13 0.446 0.370 0.435 0.451
235 densi?y,b g/cm3 0.092 0.217 0.421 0.416 0.345 0.405 0.420
Urenium coating material, kg 0.05 g.11 0.20 0,25 0.2} 0.30 0.1%
Scotch tape, kg 0.11 0.11 0.1 0.15 0.15 0.15 Q.15
Core shells ond interfece piates
Mass ot components, kg
Aluminum 5.85 1.10 .10 1.10 L1 1.10 1.10
inconel g 13.68 27.73 53.02 53.92 54.62 58.26
Refiector
Valume, f13 22,22 22,22 22,22 20.88 20.88 20.45 20.88
Minimum thickness, in. 11.5 115 11.5 1.5 115 115 1.5
Mess of components, kg
Beryliium 1155.0 1155.0 1155.0 1094.1 1094. i W077.1 1094.1
Aluminum 29.2 29,2 29.2 29.2 29.2 29.2 29.2
Excess resctivity os loaded, % 0.9 0.3 0.4 ~3 0.14 ~3 9.19
Experimental criticat mass, kg 4.35 10.8 19.8 19tz 19.9 24172  18.4
of U235 .

 

“Only one end duct wes enlarged.
bMass per unit velume of fuel region.

“Mass required for a critical system with the poison reds removed,

 

22
 

 

 

 

Fig. 10. Outer Core Shell and Peartially Assembled Beryllium Reflector
Critical Assembly.

 

UNCL ASSIFIED §
PHOTO 24437 |

of the High-Temperature

23
Fig. 11. Completed High-Tempercture Critical
Fuel Reactor.

24

 

UNCLASSIFIED
PHOTQ 2440

Assembly of the Reflector-Moderated Circuloting-

 

 

 
 

mid-plane, An analysis® of the experimental re-
sults and the differences between the critical as-
sembly ond the ART indicates that the critical
concentration of the ART will be between 4.6 and
5.4 mole % uranium, theat is, well within the {imits
of sclubility in the fuel mixture to be used in the

ART.
REFLECTOR-MODERATOR

The design of the reflector-moderator region pre-
sented several problems. Heat will be generated
in the reflector by the absorption of gamma rays
coming from the fuel and heat exchanger regions
and by the siowing down of fast fission neutrons,
Gamma rays will also result from parasitic capture
of neutrons in the structural material and the cool-
ant. One particularly strong source of hard gamma
rays will be the Inconel shell that separates the
fuel ennulus from the outer reflector. These gamma
rays will be absorbed over an appreciable volume
because the photon energy will be high and the
attenuation rather small, A lesser amount of
heating will also result from the generation of
gamma rays by neutron capture in the beryllium.
The heat generated by radiation in the various
regions of the reactor and the heat transfer from the
fuel system to the reflector and island cooling
circuits are given in Table 5,

 

SA. M. Perry, ANP Quar. Prog. Rep. Sept. 10, 1953,
CRNL-1947, p 33.

The cooling system designed for removing the
heat from the beryllium in both the island and the
reflector and from the Inconel shells is illustrated
in Fig. 1. There are 120 cooling passages in the
island and 288 in the refiecter; these passages are
0.187 in. in diameter. The dimensions of the
cocling system and the flow characteristics of the
coclant, sodium, are given in Tables 1-3. Sedium
was chosen as the coclant because of its excellent
heat transfer properties and reasonably low neutron-
capture cross section.

Experimental evidence has established the fea-
sibility of operation of a sodium-beryllium-Inconel
system if the temperoture of the system is main-
tained below 1250°F, Cycling tests have also
shown that the thermal stresses that will be set up
in the beryllium should not give serious trouble.

FUEL SYSTEMS
Core Hydrodynamics

The hydrodynamic characteristics of the reactor
core are intimately related to those of the pumps,
because the pumps must be placed in the lowest
temperature portion of the circuit, that is, just
ahead of the core inlet. This is necessary partly
because of the fairly high stresses in the impellers
and partly because of the shaft seal problem, which
is discussed in the following section on ‘‘Pumps

TABLE 5. HEAT TO BE REMOVED BY REFLECTOR AND ISLAND COOLING CIRCUITS

 

Heat to Reflector
Cooling Circuit

Heat to Isiand
Cooling Circuit

 

{(Mw) (Mw)
Radiation heating
Beryllium 1.52 0.82
Reflector B4C tile 0.48
Pressure-shell B4C tile 0.01
Control rod 0.18
Filler plotes, south head C.03
Pressure shell 0.12
Reflector outer inconel shell 0.15
North-head liner 0.03
Transfer heating
Through island core shelt 0.87
Through reflector core shell 1.73
From fuel-to-NaK heat exchanger 0.16 0.10
Total 4.04 2.16

 

25
and Expansion Tank." All the initial fayouts em-
pleyed axial-flow pumps coaxial with the istand,
but because the structural problems associated
with the long impeller overhang proved to be diffi-
cult, the hydrodynamically less desirable arrange-
ment employing centrifugal pumps was chosen
as the more practicable solution, Further it seemed
fikely that two pumps either in series or parallel
would be required if boiling of the fuel as a conse-
quence of afterheat was to be avoided in the event
that one pump failed.

Preliminary analyses of various centrifugal pump—~
core inlet configurations indicated that the high-
velocity streams frem the pump impelier would be
likely to change completely the flow pattern ob-
tained within the core. High-velocity streams are
disinclined to diffuse once they have become
separated from the wails of the pump volute, and
their momentum con carry them all the woy from the
impeiler through the plenum chamber, through the
vanes at the core inlet, and even through the core
itself to such an extent that flow separation and,
sometimes, flow reversal in the core would tend to
occur., The most promising configuration gppears
to be that in which the pump volutes discharge
tangentially into a cylindrical extension of the core
inlet, as in Fig, 2. The high swirl velocity induced
in this region gives a system relatively insensitive
to the stoppage of one pump.

From the shielding stendpoint an ideal circu-
lating-fuel reactor would have very tiny inlet and
outlet ducts to minimize both neutron leckage
through these ducts and fissioning in regiens close
to the outer surface of the reflector. While the
relations between end-duct size and shield weight
are very complex, there is o strong incentive to
minimize the end-duct size. This, coupled with
hydredynamic considerations associated with the
pump design, particuiarly the velute-discharge areaq,
and reactor physics studies, led to the choice of
the core inlet proportions and hence the basic core
fayout of Fig. 12,

The core of the reactor consists of a divergent-
convergent annular passage that is symmetrical
about the equatorial plane; that is, the converging
and diverging sections have the same shape. The
area perpendicular to the flow path ot the equator
is approximately four times the area dt the inlet or
discharge ends. The equivalent cone angle of the
divergent and convergent sections is aporoximately

28 deg (included angle),

26

In the design of the core it has been considered
necessary, in order to avoid local regions of ex-
cessive temperature, to effect a uniform circum-
ferential distribution of the flow around the infet
annulus and to assure that the fuel will traverse
the divergent section of the core without incurring
stagnation or reversal of the fluid boundary layers,
Furthermore, it is desired that the regquired flow
conditions obtain with only one pump in operation
so that the reactor can be run at some appreciakle
fraction of rated power with one of the two fuel
oumps inoperable,

It is well known that the flow of fluids in diver-
gent channels is subject tc the growth and the
eventual separation and reversal of boundary layers.
The orobability of the occurrence of these phe-
nomenc increases as the degree of divergence, as
represented by the inlet-to-cutlet area ratio, and
the rate of divergence, as represented by the equi-
valent cone angle, are increased. Flow in a diver-
gent channel is also adversely affected by uneven
distribution either circumferentially or radially ot
the inlet. Common industrial practice calls for
divergence rates to be of the order of 7 to 10 deg
included angle when preceded by several diameters
straight run of pipe. Where care has been taken
to achieve a symmetrical velocity profile with
thin boundary layers at the inlet, nonreversed flow
has been obtained in divergent channels having a
20-deg equivalent cone angle with an area ratio of
2:1. Both the degree and the rate of divergence
of the ART fuel annulus are greater than those
oreviously encountered, even under special test
conditions. In addition, the necessity for com-
pactness of the reactor rules out the possibility of
achieving even distribution and thin boundary
fayers at the infet by the most common methed,
namely, @ convergence of the channel immediately
ahead of the inlet preceded by several diameters
straight run of pipe. '

Fortunately, it is not necessarily essenticl that
the axial velocity distribution ccross the fuel
annulus be uniform; the essential requirement is
freedomn from hot spots, paorticularly at the walls.
The wall hot spot, or boundary layer heating effect,
is very much o function of the intensity of eddy
diffusivity, or turbulence, in the core. Thus, from
the standpoint of boundary layer heating, a non-
uniform velocity distribution might actually be
more desirable if the turbulence level were high
than would a more uniform velocity distribution if
the turbulence level were fow.
 

 

DT
ORNL-L.R-DWG 16043

 

 

 

  

 

 

 

 

 

 

 

 

 

 

 

- ¥oymi | O
|

Lo

<

—_—— K2

W0

: ~

il o

S o

e

M/l_

Lo

— 3

0

e

““““““““““““ ey~

 

3400 40 dOL

DISTANCE FROM EQUATOR {in.)

Fig. 12. Core Layout.

27
The practice of interposing screens across a flow
channel to achieve symmetry of the velocity profile
is widely followed, It is also well known that
screens ocross the discharge of a divergent chan-
nel tend to stabilize flow and to delay stognation
or flow reversals. In the ART fuel annulus it was
highly desirable to avoid the use of screens for
either purpose, since they would require increased
pumping power to overcome the added resistance to
flow and since they would, if used in the fissiening
zone, increase the critical fuel concentration be-
cause of their poisoning effect.

The core flow problem was therefore twofcld.
Good flow distribution was to be obtained at the
inlet, in the limited space availchle, with either
one pump or both pumps coperating; and flow through
the core was to be without stagnation or other
effects that would give local hot spots in the fiuid.
Also, these conditions were to be achieved with o
minimum of pressure loss and, preferably, without
insertion of Inconel bedies into the fissioning zone
of the core.

The desired flow distribution at the inlet was
obtained with a swirl-type heoder, which provides
space at the inlet for circulation of an excess of
fuel; that is, the volume of fuel which circulates
around the inlet is greater than that which enters
the inlet, and thus even circumferential distribution
of the flow results.

The approximate volume of the core was dictated
by fuel concentration considerations, and core
annulus radii were set at the irlet and ot the
equator, A simple two-dimensicnal flow analysis
indicated that pressure gradients resulting from
passage curvature would be small relative to those
resulting from divergence, An arbitrary cosine
curve connecting the mean radii at the inlet and
equator was therefore used as the passcge mean
line, The shape of the annulus was then deter-
mined by superimposing on the mean line a sched-
ule of cross-sectional areas such that the resulting
static pressure gradient at any peint weuld be a
function of the locel dynamic pressure,

Yon Doenhoff and Tetervin® found the function
H, in the expression

dH ‘6 dg\ /2 '
6_;__ _ ea(H—b)l: ~ (= L@’)(fi) —o(H — ar)J .
X \g dx/\T,

/

 

6A. E. von Doenhoff and N. Tetervin, Determination
of General Relations fjor the Bebhavior of Turbulent
Boundary Layers, NACA<772 {1943}.

28

to be a criterion for boundary layer separation,
where

a, b, ¢, and d are constants,

If is the shope parameter,

g is the local velocity pressure outside the
boundary layer,

x is the distence along the axis of the channel,

0 is the boundary layer momentum thickness,

Ty is the wall friction.

tf H, 0, and the term 2¢/7, are assumed to be con-
stant with x, the shape parometer con be written

where A and B are constants, P is static pressure,
and (1/4} dP/dx can be used as a criterion of
separation. Obviously (1/q) dP/dx is minimized if

it is constant throughout the divergence. However,

if this were the case, the passage weuld still be
diverging at the equator, and an undesirable dis-
continuity would result in the shape of the core
shells af this point., The divergence is therefore
modified so that it is higher at the inlet than would
be the case for

P dpP

- = a consfant,
g dx

and if is zero af the equator,

The core shape thus cobtained was tested for flow
profiles, with water as the working fluid and with a
simulated swirl-type header. The results showed
that the tangential velocities were high throughout
the core but that a small upward velocity component
prevailed clong the inner wall of the diverging
section. At all other points the axial velocity was
downward. The reason for this flow pattern is
obvious. Rotation of the fluid about the axis of the
cere creates a radial pressure gradient. Decay of
the rotational velocity component as @ result of
friction as the fluid moves axially creates an axial
pressure gradient that is positive along the inner
wall and negative along the outer. These gradients
are algebraically additive to the positive gradient

resulting from the divergence of the passage. The

net effect favors flow reversal on the inner wall.

It wos determined by test that if a solid-body-
rotation pattern (retational velocity compenent di-
rectly sroportional to radius) were used the natural
tendency for separation to occur on the outer walli,

-

 
 

 

as exhibited by irrctational flow, would be over-
come when the absolute velocity vectors at the
outside diamster of the inlet were between 15 and
20 deg off the axis. Accordingly, a set of inlet
guide vanes was designed for use with the swirl-
“type header which would set up a 20-deg solid-body
rotation. The guide vanes were designed so that
the equivalent cone angle of the intervening pas-
sages would be 10 deg, with the schedule of di-
vergence following the relation dP/dx = ag clong a
2/1 ellipitical mean line, The resulting blades had
blunt trailing edges that blocked approximately 17%
of the inlet passage area. The trailing edge area
was distributed so that the blockage occurred in
the mid-passage region, with no blockage next to
the walls. As was expected, the inlet guide vane
system gave flow reversal along the inner wall,
This reversal was eliminated by o drag ring which
biocked part of the inlet area af the blade trailing
The size ond location of the ring were
This combination of

edges.
determined by experiment.
header, core shape, and inlet guide vane system
gave flow in which the throughput component was
not stagnated or reversed of any point on either
wall. The pressure loss across the inlet guide
vane system was less than that obtained with no
guide vanes. In other words, the inlet guide vanes
recovered part of the inlet velocity head. The re-
fations between the inlet headers, inlet guide
vanes, and core are shown in Fig. 13. ‘

The flow problem is made particularly difficult by
the design having to be evaluated by experimental
tests. Vorious techniques, including pitot trav-
"erses, flow visuvalization through the use of dye
injections, and conductivity probe measurements on
sait injections, are being used. Tests that are felt
to be definitive are being carried out on the most
promising designs, both with and without vanes, in
a one-half-scale model. The volume heat source
of the reactor core is simulated with electrically

heated sulfuric acid.

While the results of the tesis made fo date are
still being studied, and they certainly pcse ques-
tions that have yet to be resclved, it does appear
that the system performs better without the inlet
guide vanes than with them. The principal problem
either with or without the vanes appears to be that
of temperature fluctuations at the wall caused by
eddying of the fuel., An experimental evaluation of
the amplitude that can be tolerated for such fluc-
tuations is under way,

i

CRNL—LR—DWG 15014

 

Fig. 13. Diagram of Core Flow System Showing
Relations Between the Inlet Headers, Inlet Guide
Vanes, and Core.

Pumps and Expansion Taonk

Various types of pumps were considered for use
in the high-temperature-liquid systems. Those seri-
ously considered included conventional centrifugal
and oaxial-flow pumps and electromagnetic and
canned-rotor pumps. Many of these pumps were
tested ot ORNL, and some performed quite suc-
cessfully.” Each type was found to have advan-
tages and disadvantages that had to be evaluated

 

7E, s. Farris, Summary of High Temperature, Liquid
Metal, Fused Salt Pump Development Work in the
ORNL-ANP Project for the Period [July 1950-jan,
1954, ORNL CF-54-8-234 (Auvg. 1954).

29
according fo the operating requirements of high-
temperature-liquid systems for aircraft instolilations.

A most important requirement of an aircraft type
of pump is that its weight be reasonable. From the
standpoint of the aircraft designer the weight
should include all the equipment required to drive
the pump; that is, if an electric motor were used,
the weight of the electrical generctor should also
be included. The importance of the weight of the
drive equipment makes the efficiency of the pump
alsc an important consideration. The combination
of these two factors eliminates electromagnetic
pumps from consideration for aircraft applications,
because the weight of an electromagnetic pump is
inherently from 100 tc 1000 times greater than that
of a centrifugal pump driven by a hydraulic motor
or an air turbine. it appecrs that a centrifugal-pump
drive-system weight of about 1 lb/hp can be ob-
tained with an air bleed-coff turbine type of system.
A further disadvantage of electromagnetic pumps is
thot they are not suitable for operation with molten
salts.

Canned-rotor pumps were also considered, but
they have the same disadvantage thet the electro-
magnetic pump has of requiring heavy electrical
generators. A further disadvantage is that the thin
sheil required between the rotor and the field
magnets constitutes a frangible diaphragm in the
system. A very large jet of fluid may be ejected
when such @ rupture occurs, and a serious fire
would result. Pumps that do not require frangible
diaphragms in the system may give trouble by
producing small lecks which would be annoying and
troublesome, but such difficulties are relatively
trivial when compared with major abrupt ruptures.
Even though the canned-rotor type of pump is suit-
able tor operation with molten salts ond even
though failure of a canned-rotor pump diaphragm
should not lead to a sericus fire with molten salts,
the large amounts of radicactivity that would be
expected in the molten salts in a full-scale reactor
would constitute o for more serious hazard than the
fire associated with sodium or NeK. Thus it ap-
pears that neither the electromagnetic nor the
canned-rofor type of pump is well suited to nuclear
aircraft application.

Quite a variety of mechanical pumps was con-
sidered, but the mixed-flow and radial-flow types
of centrifugel pump seem to be much the best
adapted to aircraft requirements. Since it is es-
sential that the reactor core and the heat ex-
changers be as compact as possible, it is neces-

30

sary to mcake use of rather high pressure drops
through these components. This in turn means that
the pump heads must be between 30 and 300 ft.
Therefore if an axial-flow type of pump were em-
ployed, it would be necessary to use multiple
stages, Flow rates of 500 to 5000 gpm will be
required, and thus relatively high shoft speeds are
essential if the impeller diameter is to be kept 1o
¢ to 12 in. [mpellers of this small size are es-
sential if the installation is to be kept reasoncbly
cempact. '

The principal problem in the design and develop-
ment of a centrifugal pump for high-temperature
liquids is the shaft seal, and quite a number of
seals were considered.?® One of the first con-
sidered was a graphite-asbestos packing placed
around the pump shait, with the gland either in the
fluid being pumped or in the gas space chove the
fluid in a sump pump. This type of seal tends to
give a considerably higher leakage rate than is
acceptable and a relatively short shoft life, since
the shaft wear is substantial, It works tolerably
well when used above the gas space in the sump
pump; but, if the seal is placed in the pumped fluid
so that there is seepage through it, oxidation of the
fluid takes place ot the cutboard end of the seal,
and high shaft wear and corrosion rates are likely
fo result,

An unusual type of seal that has received o con-
sidercble amount of attention is the frozen seal.”
This type of seal was first developed for use with
sedium, |t depends on the use of a cooled gland
arcund the shaft that is flocded with the fluid being
pumped, Friction between the pump shaft and the
frozen fluid in the gland is sufficient to melt o
very thin film at the shaft surface. A seal of this
type works well with sodium, because the shear
strength of the sodium is only of the order of 50 to
100 psi at room temperature, and aiso well with
fead, which has a shecr strength of several hundred
pounds per square inch at a temperature of arcund
200°F. The freezing temperatures of sodium and
lead can be readily oktained with a water-cooled
glond, Efforts to make this type of seal work with
Nak hove been unsuccessful because of the seal
gland having to be cooled to well below the eu-
tetic temperature (about —15°F). Unfortunately,
the hardness values of the fluoride fuels are muych
higher in the temperoture range immediately below

 

8. E. Schmitz, Trans. Am. Soc. Mech, Engrs. 71,
635 (1949).

 
 

their melting points thon the hordness value for
sodivm. Frozen seal experiments have been made
with many fluoride mixtures, including many glassy
melfs with large percentages of beryllium fluoride,
but in all instances serious cutting of the shaft
occurred within a few hundred hours. In addition,
large amounts of power were required for contin-
vous cperation, and very lorge amounts of torque
were necessary in order to breck the frozen seal
during startup of the pump; pumps normally re-
guiring only 3- to 5-hp motors were found to require
30- to 50-hp motors to be started.

The face type of seal used widely in automobile
water pumps, refrigerant pumps, end domestic water
pumps appears to be a promising solution to the
seal problem. [n most applications it is flooded
with the pumped fluid during operation. Its very
low leakage rate and long fife depend on the mating
faces of the seal being finished to essentially
eptically flat surfaces. The liguid tends to fill the
space befween the two seal surfaces ond to form a
meniscus between the edges of the seal faces on
the gas side. The surface tension in the meniscus
across this very narrow gap gives a pressure in the
fluid between the seal faces that is sufficient to
hold the surfaces aparf, Therefore the surfaces do
not contact each other, but, rather, they shear the
fluid film between them. Thus the seal surfaces
operate under ideal |ubrication conditions. Seals
of this type have opercted for years with no meas-
urcbfe wear. They are relatively insensitive to
starts and stops if the seal-face pressure is kept
fow, It is important that they be adequately cocled
and that the product of the pressure in pounds per
square inch on the seal face ond the rubbing ve-
locity in feet per minute not be excessive. Values
~as high as 100,000 for this pressure-velocity factor
have been reported, with the parts giving very
satisfactory service life, [t is essential that the
mating surfaces in a seal of this fype be compatible
from the standpoint of boundary lubrication, or else
they moy be scored during storts and stops, It is
also very impertant that this type of seal be
mounted on a shaft that runs true, with a minimum
of vibration. This means that the shaft must be
well balanced with the impeller and bearing as-
sembly, that the radial lcoseness in the bearings
must be small, and that the seal must be mounted
in such @ way that it is both concentric and square
with the axis of the shaft, Also essential is
flexibility in the seal mounting, which can be
readily obtained either with a corrugated diaphragm

or a bellows, the latter being the more commonly
used, Since graphite is a porous material, the
graphite washers used in seals of this type are
ordinarily impregnated with materials suck as a
plastic, lead, babbit, silver, ete. For high-temper-
ature use in the ART fuel pumps the seal will
operate with inert gas on one side and a flood of
oil on the other, rather than fuel, to prevert con-
tomination of the fuel.

The bearing problem has much in common with
the seal problem for high-temperature-fluid pumps,
In all instances the fluids are very corrosive to
most materials, and therefore only a few materials,
such as graphite, certain iron-chrominum-nickel al-
lovs, and cemented carbides, can be used. Further,
the fluids pumped will remove any adsorbed films
such as sulfides, phosphides, etc., that would
tend to alleviate boundary-layer [ubrication con-
ditions, The viscosity of sodium is about one-
fiftieth that of water, while the viscosity of the
fluoride fuel mixture is about the same as that of
water. Thus neither of these fluids serves as o
really good lubricant. While they have the advan-
tage of wetting the surfaces of iron-chromium-nickel
alloys very effectively, they tend to strip off the
protective films that are formed under ordinary con-
ditions in most types of petroleum- or other nydro-
carbon-lubricated bearings. It is evident that
bearings designed to operate in molten metal or
molten salt must be lightly loaded and carefully
aligned, .

The investigations have indicated that the bear-
ing and seal should be placed in @ cool zone above
the pump impeller with a heat dam between them
and the fluid being pumped. This arrangement
makes it possible to use conventional bearings
and seals. Pumps of this type have proved te be
quite satisfactory and have the advantage of being
relatively insensitive to the type of fluid being
pumped; that is, the same type of pump can be used
for sodium, NaK, the fivoride fuel mixture, lead,
sodium hydroxide, or other fluids,

Two fuel pumps and two sodium pumps will be
located at the top of the reactor, These pumps will
be similar, but the fuel pumps will have the larger
flow capacity. The fuel pumps will also include
xenon-removal systems in which most of the xenon
and krypton and prebably some of the other gaseous
fission-preduct poisons will be removed from the
fluoride fuel mixture by scrubbing with helium, The
fuel will enter the xenon-removal system from the
eye of the pump and will pass up the center of the

31
shoft and inte the mixing chamber. In the mixing
chamber the fuel wiil spray through a helium at-
mosphere and impinge on the wall of the chamber,
The resulting mixture will be very foamy and wili
have a latge gas interface. The helium will enter
the system just below the shaft seal and flow down
the annulus around the shaft, through the upper
slinger vanes, and into the mixing chamber, The
fuel-helium mixture will then be pumped into the
fuel expansion volume, where the helium centaining
the xenon, krypton, and other fission-product gases
will be removed via the off-gas system. The circuit
wiil be completed when the fuel leaves the expan-
sion volume by grovity flow intc the centrifuge,
where any entrained gas will be removed and then
be returned to the expansion volume. The fuel will
be pumped through the centrifuge heles and will
re-enter the main fuel system on the discharge side
of the fuel pump. The system is illustrated in

Fig. 14.

Experimental results obtained with a single pump
operating in a test lcop showed that the fuel level
in the expansion chamber and the rate of bleed flow
affect the ability of the centrifuge to prevent helium
bubbles from entering the main fuel system. The
system operates very well with a fuel level of ¥ in.
in the expansion chamber and a bleed flow of up to
13 gpm. With a fuel level of 3 in. in the expansion
chamber the system can be operated with a bleed
flow of 25 gpm. The design value for the bypass
tlow rate is 12 gpm. Similar tests on g full-scale
aluminum model of the fuel pump—expansion tank
region have given similar results,

A number of other special features have been in-
cluded in the pump design for adaptation to the
full-scale reactor shield. The pump has been de-
signed so that it can be removed or instailed as a
subassembly, with the impelier, shaft, seal, and
bearing comprising a single compact unit. This
assembly will fit into a cylindrical casing welded
to the top of the reactor pressure shell. A 3-in.
tayer of gemma-ray shielding just above a l/z-in.,
layer of zirconium oxide around the lower part of
the impeller shaft will be at the same level as the
reactor gamma-ray shield just outside the pressure
shell. The space between the bearings will be
filled with oil to avoid a gap in the neutron shield.
The pumps will be powered by hydroulic drive units
in order to provide good speed centrol, along with
compact, reliable metors.

32

Fill-and-Drain System {Including Enricher)

The system designed for filling the reactor with
the barren fuel-carrier sait NaZrF, for adding the
enriched uranium-bearing fuel component Na,UF ,
and for draining the reacter is described schemat-
ically in Fig. 15. For the initial fiiling operation
the fuel carrier will be added to the fill-and-drain
tank at a temperature of about 1200°F and will be
pressurized into the reactor @ number of times at
this temperature. The reactor cond the fill-and-
drain tank will have been preheated by cperation
of their respective NaK systems. Approximately
60% of the uranium-bearing component of the fuel
will then be added from a simple melt pet by gas
pressure transfer. The remaining uranium-bearing
material required to achieve criticality will be
added in small increments by using the enricher
device shown in Fig. 15.

The fuel fill-and-drain system is designed to
permit the transfer of fuei to and from the reactor
after the reactor system has been completely
assembled. A heating system is included to
maintain the fuel at a temperature above its
melting point. This same system aiso serves to
cool the fuel by removing decay heat from fuel
drained after the reactor has been operated at
power.

The initial fuel charge is to be admitted to the
fuel fili-and-drain tank through a fill nozzle
provided at the top of the tank. This initial filling
operation will be carried out by persennel who will
be inside the pressure vessel ot the fill-and-drain
tank. After the addition of the barren salt to the
{ill-and-drain tank, the system will deliver this
salt into the reactor for initial testing. The barren
saft will then be drained for the addition of the
enriching mixture. The system will then be used
to fill or partly fill and drain the reactor ¢ number
of times for mixing the salt and enriching mixture.
This enriching and mixing cycle may have to be
repeated a great many times. Fuel flow into the
reactor will be by displacement with helium, and
flow back to the fill-and-drain tank will be by
gravity. The helium displaced from the reactor
duting fuel addition will be discharged through
the off-gas system. Helivm displaced from the
drain tank as the fuel is drained will flow through
part of the off-gas piping intc the top of the
reacter,

The fili-and-drain tank will also receive the
fuel charge from the reacter ot any time during
SECHET
ORNL~LR-DWG 16015

EXPANSION VOLUME

 

 

 

 

    
     
    
 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

H— SLINGER IMPELLER

 

 

 

 

P
]

EXPANSICN
VOLUME

[—MIXING CHAMBIZR

 

 

 

 

 

 

 

_}—SEAL PLATE

 

 

— ]

 

]///—_

 

CENTRIFUGE SEAL VANE

 

 

 

 

 

 

 

 

BAFFLE ——} —— CENTRIFUGE HOLZ

 

 

 

 

 

 

PUMP N\
piscHaRrGe | FUEL PUMP X/ /

 

 

 

 

 

 

 

 

 

 

 

 

 

/ PUMP SUCTION

|
|

Fig. 14. Section Through Fuel Pump~Expansion Tank Region Showing Xenon-Removal System of
the Fuel Pump.

 

33
SECTPT
ORNL-LR-[CWG {6046

 

HELIUM ____,[ J__,

OVERFLOW LINE

 
    
    

 

 

 

 

i Ps ¥
- VAPOR b
ENRICHER T

BREAK POINTS V" -

ey

HE..IUVI

TO POSTPOWER

o FILL - AND - DRAIN SAMPLER
R PRtP(,Wr_R TANK TO FUEL PROCESSING

SAMELER SYSTEM TANK

" TANK FOR ADDITION OF
CARRIER AND INITIAL
BATGH OF CONGENTRATE

Fig. 15. Schematic Diegrom of Fuel Fill-and-
Drain System, Including Enricher.

the operation when an emergency requires that
the fuel be drained. Under these conditions, heat
will have to be removed from the fuel ot o rate
equal to the fission-product-decay heat-liberation
rate.

Finally, the fuel fill-and-drain system will
receive the fuel from the reactor at the completion
of the test and will hold it as long as decay heat
requires cooling of the fuel. It will then be dis-
placed by helium pressure into the fuel recovery
tank for removal. No provision has been made for
transferring the fuel back frem the recovery tank
to the fill-and-drcin tank or for cocling the fuel
in the recovery tank.

After the preliminary design studies were com-
pleted, it was decided that the fill-and-drain tank
should be designed with two independent sets of
cooling tubes, each designed to permeste com-
pletely the fuel veclume. This arrahgement will
permit the tank to be cooled even though some
component of one of the two cooling systems, such
as a pump, radiator, or pipe, should fail. The

34

design heat load on the tank is 1.75 Mw. The
theoretical decay heat ot the instant of shutdown
in a 60-Mw reactor is given as about 3.6 Mw, but
only a 1.75-Mw cocling system is required, because
the fuel will not be drained into the drain tank
until the fission heat is negligible and untii the
fuel has been cocled to below 1200°F. The drain
valves will not be opened until at least 8 sec
after the contre! rod has been fully inserted. The
design is based upon the assumption that such
emergency fuel drainage would ecccur after contin-
vous operation at 60 Mw for one month. The heat
capacities of the fuel and mechanical equipment
are great enough tc absorb the decay heat above
1.75 Mw for the few minutes during which the
decay heat generation would exceed this rate,

The fuel piping between the reactor and the
fili-and-drain tank is designed to drain all the
fuel from the reactor under gravity flow conditions
in not more than 3 min. [t is calculated that this
condition will be met even though one of the two
drain valves cannot be opened.

The requirement for a completely reliable cocling
system seems to be met by the use of bundles of
cooling tubes inserted into the tank. In order to
provide cooling for all metal surfaces in contact
with the fuel, the tank and the standpipe to the
reactor are both jacketed. The coolant to be used
for this service will be NaK (56-44 wt %). To
provide two independent sets of cooling passages,
each of which permeates the entire fuel volume, a
cylindrical tank was selected, with
each end serving as a tube sheet. Horizontai
decks of U-tubes arranged on each tube sheet in
a square-hcle pattern provide alternate layers
that originate at opposite ends of the rank (Fig.
16}. Design calculations indicate that fellowing
an emergency drainage of the fuel, the cooling
system will, even with only one circuit functioning,
limit the fuel temperature in the tank to 1800°F.
When both coolant circuits function, the maximum
temperature at any point in the fill-and-drain tank
will be limited to 1600°F.

The jackets around the cylindrical shell of the
tank and around the drain line, however, are not
served by two separate systems. Coolant flow
through the tank jocket wiil be in parallet with
coolant flow in cne set of U-tubes., The NaK will
enter the jocket around one half of the tank
circumference and will return to the coolant
channe! cround the other half of the tube sheet
periphery.  Flow guides are placed along the

horizontal,

 
 

 

CEONFBENRAL
ORNL—~LR-~DWG 16!26

Nak SUPPLY -1 j,
{HHOG°F)

   

 

 
    
 

¥ FUEL DRAIN
¢ ° LINE

.
“~--INNER CYLINGER
{He BLANKET AT 5 psi }

-~ BAFFLES { FOR SUPPORT

- NaK SUPPLY —2 AND FLOW DIVISION)

  

L} SUPPORT CYLINDER
1 (5000 1b UPLOAD}

Fig. 16. Fuel
System.

Fill-end-Drein  Tank Cooling

annular jacket to distribute the coclant flow as
required.

The two sets of U-tubes ore arranged so that
one set cools the fuel zone just inside the cylin-
drical tank wall more thoroughly than the clternate
set, and the coolant supply to the jacket is from
the alternate system, which is least effective in
cooling the tank wall. At one end of the tank,
where the fuel shell and tube sheet intersect, a
triangular “filler ring is placed to displace fuel
from this corner zone, where coocling would be
poor if one of the two coolant circuits failed, At
this end of the tank, failure of the coolant circuit
would interrupt cooling behind the tube sheet and
in the jacket. At the other end of the tank, coeling
would continue either in the jacket or behind the
tube sheet, and no filler ring is required.

Further details of the operation of this system
are given in the section “Operation of the ART"
and on the flowsheets in Appendix A.

Sampling System

Samples of the barren molten salt will be token

immediately aofter ‘‘shakedown’’ operation of the

ART by pressurization of the melten salt through
a heated Inconel tube from the fill-and-drain tank
into a detachable sample receiver. Additional
somples will be taken during enrichment in the
same manner fo defermine the homogeneity of
mixing and the concentration of uranium in the
mixture,

Samples of the molten fuel mixture will also be
taken after nuclear power operation. These samples
will aiso be transferred through hected Inconel
lines into sample receivers, but, because of the
high level of radicactivity in the fuel, the operation
is to be accomplished by remote control, and the
lines and receivers must be shielded. To prevent
premature activation of the sampling valve, a
frangible disk valve will be included in each
sample transfer line.

A total of five postpower samples may be taken.
The five receivers wiil be encased in the same
lead shielding urit, which will be removed after
the entire ART opercation is completed.

Fuel Recovery System

After completion of ART operation, the fuel will
be drained into the fill-and-drain tank and held
there for a period of several days to allow fission-
product decay and the consequent lowering of
decay heat production. The fuel will then be
transferred into the fuel recovery tank for delivery
to the recovery and reprocessing facility.

The fuel recovery tank will consist of four,
interconnected, approximately 9-ft lengths of 8-in.-
IPS Inconel pipe furnished with elecirical heaters.
The assembly is designed for cooling by radiation
ond natural convection to the atmosphere. It will
be shielded with approximately 10 in. of lead.
The molten fuel will be transferred to the tank
through a heated Inconel line containing a frengible
disk valve and an open bismuth valve. When the
transfer is completed, the bismuth valve will be
closed and both the transfer line and the helium
line will be severed, as will all heater and thermo-
couple leads, The assembly will then be lifted
out of the cell and loaded on a truck for delivery
to the recovery facility. The assembly will
include o transfer line and the other necessary
service connections for use in the course of
recovery operations.

HEAT EXCHANGERS AND HEAT DUMPS

The spherical-shell fuel-to-NoK heat exchonger,
which mokes possible the compact layout of the
reactor heat exchanger assembly, is based on the

35
use of tube bundies curved in such a way that the
tube spacing will be uniform, irrespective of
lotitude.® The individual tube bundles terminate
in header drums. This orrangement focilitates
assembly because small tube-to-header bundles
can be assembled, made leaktight, and inspected
much more easily than can one large unit. Fur-
thermore, these tube bundles give a rugged,
flexible construction that is admirably suited to
service in which large omounts of differential
thermal expansion must be expected, The config-
uration of the heat exchanger tube bundles is
iHustrated in Fig. 17.

The heat removed from the fuel by the NaK wili
be transferred to air in the NaK-to-agir radiators,
which will be installed in ¢ hect-dump system
designed to simulate the turbojet engines of the
full-scale aircraft in a number of important respects,
such as thermal inertia, NaX holdup, and basic
fabricational methods. The round-tube and plate-
fin radiator cores are fabricated of type 310
stainless-steel-clad copper fins spaced 15 per
inch and mounted on %6-En.=OD tnconel tubes
placed en 2/3=-irx., square cenfers. 1 he tubes are
welded ond brazed into round header drums. The
individual radictor cores consist of twe halves
15 by 30 in. The fin matrix depth in the air flow
direction is 5.33 in. A typical radiator is illustrated
in Fig. 18.

The basic reguirement of the heat-dump system
is to provide heat-dump capacity equivalent to
40 Mw of heat with ¢ mecn temperature level of
1300°F in the NaK system. The NaK will be
circuiated through eight separate systems. Four
will constitute the main heat-dump system, two
witl serve the reflector-moderator cooling system,
and two will serve the fuel fitl-and-drain tank.

in the main system a group of four NaK fill-and-
drain tanks will be used. The tanks will be pres-
surized to force the NaK into the main cooling
circuit. The 12 tube bundles of the fuel-to-NaK
heat exchanger will be manifolded in four groups
of three each. The NaK will flow from these tube
bundles out to the radiators, which will be arranged
in four vertical banks with four radiotors in each
bank. The NaK will flow upward through ihe
radiator bank to the pumps. A small NaK bypass
fiow around the radiators will pass threugh o
cold trap in order to maintain o low oxygen concen-

 

%A. P. Frags and M. E. LaVerne, feat Exchanger
Design Charts, ORNL-133C (Dec. 7, 1952).

36

tration in the NaK. (For details of the NaK systems
see Appendix A, Flow Diagram 6.}

The two independent reflector-moderator heat-
dump systems (referred tc on Flow Diagram 6,
Appendix A, as the Auxiliary System) will be
essentially similar to the main NaK system except
that their combined capacity will be about one-
third thot of one of the four circuits of the main
heat-dump system. The NaK will be circulated to
both the Na-to-NaK heat exchengers in the top of
the reactor, where the two NaK streems will pick
up from the sodium the heat generated in the island
and the reflector. The twe NaK streams will pass
to small NaK-to-gir radiators, where they will be
cooled ond returned to their respective pump
suctions. A bypass cold trap will be inciuded in

each system, as in the main NoK systems, while

a single fill-and-drain tank will serve both systems.

Four axial-flow blowers will force 300,000 cfm
of air {which expands to 6.7 x 10° cfm when heated
to 750°F) through the radicfors and out through
a 10-fi-dia dischorge stack 78 ft high. Since the
axial-flow blowers will stafl and surge if throttled,

contrel will be accomplished threcugh bypassing @

portion of the air around the radiators. The heat-
dump rate will be modulated by varying the amount
of air bypassed through a set of conirollable
jouvers mounted in such & way as to bleed air
from the plenum chamber between the blowers ond
the radiators. The arrangement will include louvers
te block off the air passage to the radiators and
louvers in the bypass duct; one set will be
opening while the other is closing. This arrange-
ment should give good centrol of the heat load
from zero to 110% of the design load. Since each
biower will be driven with an a-¢ motor independ-
ently of the others, the heat-dump capacity can
also be increased in increments of 25% from zero
to full load by changing the number of blowers
used.

Heat barriers mounied on both sides of the ra-
diators will be required in order to minimize heat

losses during wormup operatiens. Warmup will be”

accomplished by energizing the NaK pumps and
driving them ot part or full speed. As a result of
fluid frictional losses approximately 400 hp must
he put into the pumps in the NaK circuits and will
appear as heat in the fluid pumped. A mechanical
power input of 400 hp to the NaK pumps will
sroduce a heat input in the Nall system of approxi-
mately 300 kw. This should be encugh to heat
the system quite satisfacterily if the radiator
O
o
@
o
O
bed
O
-
O
I
'
p 4
o
o
Yo

 

 

 

 

 

 

R PV,

 

MNaK Heot Exchanger Channel.

af Ge

in Fuel

Mode!l of Ma

. 7.

g

@

F

 

37
| UNCLASSIFIED
1

 

Fig. 18. Prototype ART NaK.tc-Air Radiator.

cores are blanketed to prevent excessive heat
losses. However, electrical heaters will be avail-
able on the NaK lines so that the NaK pumps can
be run at one-half speed to reduce stresses and
weor during zero- ond low-power operation. Rela-
tively simple sheet stainless steel doors with
3.0 in. of thermal insulation will, when closed
over both faces of a radiator (100-ft2 inlet-face
area) filled with 1100°F NaK, reduce the heat loss
there to about 30 kw.

The heat appearing in the reflector-moderator
will be cbout 3.5% of the reactor power oufput.
The cooling circuit will also remove heat from the
core shells ond the pressure shell, and therefore
the total amount of heat to be removed from this
cooling circuit will be ebout 10% of the reactor
output. This must be removed at a mean NaK
circuit temperature of about 1050°F. Four ra-
diators that have inlet faces 2 x 2.5 ft each and
the same basic geometry as that ysed for the main
heat dumps wiil be employed. These radiators
will be equipped with louvers and heat-barrier
doors like those used on the main NaK radiators
and will be supplied with air from the same air
duct {see Flow Diagram 6, Appendix A).

38

The two independent heat-dump systems for
heating and cooling the fuel fill-and-drain tank
(referred to on Flow Diagram 6, Appendix A, as
the Special Systems) will also be essentially
similar to the main NaK system. The two NaK-to-
air radiators in these systems will be served by
individual air blowers, When the systems are to
be used for adding heat to the fuel in the fill-and-
drain tank to preheat the tank for filling or to
maintain the fuel above its melting temperature,
they will be capable of heating the tank to 1250°F.
The systems will also be ready at any time to
remove fission-product-decay heat from the fuel.

OFF-GAS SYSTEM

The fission-product gases which will be purged
from the fuel in the fuel expaonsion tank are to be
centinvously removed from the fuel system to a
water-cooled charcoal adsorber, in which the gases
will, in effect, be “‘held up’’ « sufficiently fong
period of time for safe discharge to the stack. In
the fuel expansion tank the gases will be diluted
with 1000 to 5000 liters of helium per day and will
be bled first through a water-cooled vapor trap (te
remove ZrF, vapor) and then outside the reactor
and cell through an empty pipe into a charcoal-
filled pipe in a water-filled tank. The empty pipe
was provided to permit some decay of the gaseous
activity so that when it is subsequently adsorbed
by the charcoal the resuitant heat can be satisfac-
torily transferred to the water surrounding the
charcoal-filled pipe. There will be sufficient
charcoal to assure a 48-hr holdup of K¢88, which
will be the only gaseous fission product with
significant activity after passage through the
charcoal. Experimental work has indicated that
64 13 of charcoal will be required. Two parallel
charcoal adsorber systems are provided.

Provisions have also been made for bleeding
the cell atmosphere through an auxiliary charcoal
bed in the event that a leak occurs in the reactor
off-gas system within the cell. The charcoal bed
in this auxiliary vent system will be bypassed
during normal operation for disposing of insfrument
bleed gases.

Four bypass loops will be provided in the two
off-gas systems, one before and one after the
charcoal section of each system, so that the
activity in the isolated gas samples can be counted.
Further, the vent lines to the stack will be mon-
itored to determine whether the gas may be safely

 
 

 

discharged. The system will operate at o pres-
sure slightly above atmospheric and will not
require pumps.

Except for the equipment within the reacter cell
and in the water tank (containing the empty and
the charcoal-filled pipes), which will be buried
in the ground, all components of the off-gas system
are to be housed in a special building provided
for that purpose at the southwest corner of the
ART building. A fundamental design criterion
is that access to this “‘off-gas shack’ should not
be limited by radicactivity from any other part
of the system but that the shack should be isclated
in the event of a leck of gaseous activity therein.

In the design of the off-gas system the calcu-
lations that were made!®~12 were strictly theoret-
ical, but they were strengthened by the operating
experience which had been obicined on a similar
adsorber system used with the HRE. In particular,
flow vs pressure-drop data and flow vs breakthreugh
times in charcoal adsorbers were obtained from
HRE experience.'S FExperimental data'4 on the
adsorptivity of charcoal indicate that the design
of the ART adsorber is conservative. Details of
the system are presented on Flow Diegrem 3 in

Appendix A.

AIRCRAFT-TYPE SHIELD

The mest promising of the several shielding
arrangements that were considered for the ART
seems to be the one which is functionally the same
as the arrangement for an aircraft requiring a unit
shield — a shield designed to give ~10 r/hr at
50 ft from the center of the reactor. Such a shield
is not far from being both the lightest and the most
compact that has been devised. It will make use
of noncritical moterials that are in good supply,
and it will provide useful performance data on the
effects on the radiation dose fevels of the release
of delayed neutrons and decay gammas in the heat
exchanger, the generation of secondary gammas
throughout the shield, etc. While the complication

 

Wy, B. Cottrell et i, Aircraft Reactor Test Hazards
Summary Report, ORNL.-1835, p 24 (Jan. 19, 1955).

”Ca S. Burtnette, Fission Product Heating in Off-Gas
System of the ART, ORNL CF-55-3-191 (March 28,
1955).

12, B. Andersen, Adsorption Holdup of Radicactive
Krypton and Xenon, ORNL CF-55-8103 (Aug. 154, 1955).

mh Spiewak, Use of HRE Charcoal Adsorbers in the
HRT, ORNL CF-54-7-26 (July 8, 1954).

Yy, E. Browning and C. C. Bolta, ANP Quar. Prog.
Rep. March 10, 1956, ORNL.-2061, p 193.

~ spherical-shell

of detailed instrumentation within the shield does
not appear to be warranted, it will be extremely
worth while to obtain radiaticn-dose-level data at
representative points around the periphery of the
shield, particulorly in the vicinity of the ducts
and of the pump and the expansion tank region.

The shield for the reactor will have some chorac-
teristics that will be peculior to this particuler
reactor configurction. The thick reflector was
selected on the basis of shielding considerations.
Use of o thick reflector is based on twe major
reasons: a reflector about 11 in. thick followed
by o layer of boron-beoring material will attenucte
the neutron flux to the point that the secondary
gamma flux can be reduced to a value about equal
to that of the core gamma radiation; it will also
reduce the neutron leakage flux from the reflector
into the heat exchanger tc about the level of that
from the delayed neutrons which will appear in
the heat exchanger from the circuleting fuel., An
additional advantage of the thick reflector is that
99.8% of the energy developed in the core will
appear as heat in the high-temperature zone within
the pressure shell. This means that very little
of the energy produced by the reactor will have
to be disposed of with a perasitic cocling system
at a low-temperature level. The material in the
intermediate heat exchanger is
about 70% as effective as water for the removal
of fast neutrons; so it too is of value from the
shielding standpcint. The delayed neutrors from
the circulating fue! in the heat exchonger region
might appear to pose a serious handicap. However,
they will have an attenuation length much shorter
than the corresponding attenuation length for
radiation from the core. Thus, from the ocuter
surface of the shield, the intermediate heat ex-
changer will appear as a much less intense source
of neutrons than the more deeply buried reactor
core. The fission-product-decay gammas from the
heat exchanger will be abeut equolly as important
as the secondary gammas from the beryllium and
the reflector shell.

Thermal insvlation 0.5 in. thick will separate
the hot reactor pressure shell from the gamma
shield, which will be o loyer of lead 4.3 in. thick.
The lead, in turn, will be surrounded by a 32-in.-
thick region of borated water. The slightly pres-
surized water shield will be contgined in an
aluminum tonk. Cooling of the lead shield will
be effecied by circulation of woter through coils
embedded in the lead. The borated water shield

39
will also be ccoled by water circulated in coils
submerged in the borated water and by thermal
convection of the atmesphere in the reactor as-
sembly cell.

An opening will be made at the bottom of the
shield for filling and draining of the reactor. The
fuel fill-and-drain tank will be shielded with 10 in.
of lead to reduce the dose to 1 r/hr at the shield
surface one week after full-power operation. The
resulting shield weight for a 13-f° capacity tank
will be about 30 tons.

CONTROLS AND INSTRUMENTATION

The early effort by ORNL personnel to develop
the circulating-fuel type of aircroft reactor was
motivated in part by o desirable control feature
of such reactors — the inherent stability of the
reacfor ot design point that results from the neg-
ative fuel and over-all temperature coefficients of
reactivity. in a power plant with this characteristic
the nuclear power source will be a slove to the
turbojet load with but a minimum of external control
devices.

This predicted master-slave relation between
the foad and the power source was verified by the
ARE. Work with an electronic simulator indicaftes
that the ART should behave in essentially the
same way. Controlwise, the power plant consists
of the nuclear source, the heat dump (in the case
of the ART), and the coupiing between source
and sink (the NaK circuit). Control at design
point can be effected to some extent by nuclear
means at the reactor, by changing the coupling
(i.e., changing the NaK flow) or by changing the
load (i.e., the heat dump from the NaK radiators).

For the ART ot design point the regulating rod
wifl be used mainly for adjusting the reactor mean
fuel temperature., In particular, an upper tempera-
ture limit will cause the regulating rod to insert,
and therefore the fuel outlet temperature should
not appreciably exceed 1600°F, even in transients.
This limit will override any normal demand for
rod withdrawal. Furthermore o low NagK ocutlet
tempercture from the heat-dump radicters will
avtomatically decrease the heat lcad to keep the
lowest NaK temperature of the system ot no less
than 1070°F. This lower temperature limit will
override all other demands for power,

Control of the ART falls in thiee different
categories of operation: stertup, eoperation be-
tween startup and appreciable power (about 10%
of design point), and operation in the range from

40

10% to full power. For the second and third of
these categories the nature of the reactor and
power plant is so different from that of conven-
tional high-flux reactors that control must be
based on inherent characteristics of the reactor
to a large extent rather than on conventional
reactor-control practice. Control at startup utilizes,
in principle, old reactor-control practice with
short-period  ‘“’scrams’’ that are conventional,
Experimentation will take place primarily in the
startup and design-point regions. In the inter-
mediate region between these two, little testing
will be done.

Fission chambers and compensated ion chambers
witl be locoted beneath the reactor shell just
outside the lead region. The region around the
fill-and-drain pipe will be fililed with moderator
material through which cylindrical holes for these
chambers will run radially out on lines which
intersect at a point below the center of the reactor,
The chamber sensitivities will be adequate for
the entire range of nuclear operations.

The contrel system is designed to provide auto-
matic corrective action for emergencies requiring
action too rapid to permit operator deliberation.
Automatic  interfocks will prevent inadvertent
dangercus operation, with minimum operator limita-
tion. Operation in the design power range will
be independent of nuclear instrumentation during
power transients. Three classes of emergencies
have been provided for.

Class I. — Any of the following events will cause
the rod to be automatically and completely inserted
at a rate of 1% Ak/k in 8 sec:

I. stoppage of either sedium pump,

2. stoppage of either fuel pump,

3. drop in fuel or sodium liquid levels in the ex-
pansion tanks when pumps are operating af
design peint,

4. failure of the oil system to the sodium or fuel
pumps,

5. leakage of fuel into NaK or NaK into fuel,

6. failure of commercial power or locally gen-
erated power.

At the same time one-half the blowers will be shut

off to reduce the heat load. The load will then

be further reduced by the radiator shutter auto-
matically closing in response to a 1070°F low-
signal for the NoK-te-air radiator outlet

Dumping of the fuel will not be

automatic but will probably be initicted by the

operdator.

limit
temperature.

 

 
 

Class I, — Any of the following events will
cause the rod to be automatically inserted ot o
rate of 1% Ak/k in 8 sec:

1. stoppage of any NaK cireuit,

2. leakage of NaK to the atmosphere,

3. trouble in the off-gas system (to be defined),
4. maximum fuel temperature greater than 1650°F,
5. maximum

1300°F,

6. maximum sodium temperature greater than
1250°F when either set of sodium-to-NaK
radiator lcuvers is wide open,

7. reactor on positive period of less than 3 sec,

8. fill-and-drain tank mecan temperature grecter
than 1300°F or less than 1100°F,

9. failure of the oil system to any NaK gump,

10. rod-drive trouble.

Class I, —~ |t is planned that the operator will

be warned of any of the following events, but no
automatic corrective action will take place:

sodium temperature greater than

1. leakage of sodium into fuel or fuel into sedium,
2. lowering of the water level in the outer cell,

3. excessive radiation level in the cell as deter-
mined by monitors,

excessive humidity in the cell,

excessive rod temperature,

excessive chamber temperature,

. oxygen in the cell.

Lists of the instruments for the ART have been
prepared that give the foliowing information for
each required measurement: type of instrument
pickup, location of pickup, type of presentation,
reading range, and accuracy. |he
necessify for using an instrument was determined
by whether it would be required for the safe and
orderly conduct of the test, for providing sufficient
information for evaluation of test results, and for
previding not otherwise available,
Five stations have been provided in the ART
building at which instruments will be read: the
control room, the information room, the auxiliary
equipment panel, the vent house, and the temporary
panel for fuel sampling and recovery.
~ All instruments pertinent to the nuclear perform-
ance of the reactor, as well as to the control of
equipment which coffects nuclear
performance, are located in the control room.
The instruments required for determination of
reactor power and heat exchanger and radiator
performance are located in the information room.
In general, operating instruments for auxiliary
systems, such as water, hydraulic fluid, lubricating

SO A

location,

informetion

all process

oil, gas, etc., as well as for the NaK system and
all pumps, will be confined to the auxiliery equip-
ment panel. A few pertinent instruments, as well
as a number of alarms, are duplicated in both the
control room and the cuxiliary equipment panel.
With a few minor exceptions, all heater controls
and associated femperature instruments are located
in the basement on the cuxiliary equipment panel.
Off-gos system temperatures will be recorded in
the vent house to avcid lines being run to the
information room. The temporary panel (for fuel
sempling ond recovery) will be located on the main
floor just outside the cell and will be connected by
temporary lines to equipment within the cell as
required to effect the fuel sampling and fuel
recovery operations.

AUXILIARY SYSTEMS
Heliom Supply System

Helium is required in the ART principaily be-
cause of a need for a flushing gas for removing
fission gases, for an inert atmosphere over the
tuet,
medium for forcing fuel and NaK to and from their
respective fill-and-drain tanks inte the recctor
system. Since no two of these three uses of helium
will be concurrent, the helium consumption rate
will be tow. Therefore the helium is to be supplied
from 12 cylinders manifolded into two banks by a
high-pressure manifeld with isolating valves and
pressure-reducing valves for each bank. This
will allow use from either of the two banks while
the depleted bank is being reploced. (For details
of this system see Flow Diagram 9, Appendix A.)

sodivm, and NaK, and for o pressurizing

Nitrogen Supply System

Nitrogen will be required for filling the recctor
cell and for operating the pneumatic instruments.
The nitrogen atmosphere in the cell is a safety
measure. In the event of a simultaneous sodium
and water leak from the reactor into the cell, it
is important thot the oxygen concentration of the
ambient atmosphere be kept low encugh for no
detonation of the hydrogen-oxygen mixture to
occur. fhe lower combustibility limit for hycrogen-
nitrogen-oxygen mixtures is obtained at 5% oxygen.
To ensure a substontial margin of safety, it has
been decided that the oxygen concentration in
the reactor cell will be kept to less than 1%.

The oxygen will be removed ‘initially from the
cefl by the cell pressure being reduced with o

41
vacuum pump to a maximum of 0.74 psia and the
cell being repressurized with dry nitrogen fto
atmospheric pressure. The dry nitrogen bleed from
the preumatic instruments will act as a continucus
purge, which will keep the oxygen content of the
nitrogen low. The bleed flow is estimated to be
a maximum of 7 scfm. (For details of this system
see Fiow Diagram 10, Appendix A.)

A permanent storage tank with a capacity of
10,500 scf ot 1000 psi and a pressure-reducing
station with two reducing valves that are available
at the site will comprise an adequate system for
supplying the instruments with the 7 cfm required
during the test. A trailer will be used to fill the
reactor cell initially.

Electrical Power System, Distribution,
and Auxiliery Equipment

The electrical power for the ART will be sup-
plied by two separate scurces. One will be a
commercial (TVA) source and the other will be
a set of diesel-driven generators. In case of a
power failure for any reason, no effort will be made
to continue to operate the reactor at power; that
is, a power failure or an equipment failure will
lead to an orderly shutdown of the reactor. The
main power loads are:

1. the two fuel pumps,

2. the two sedium pumps,

3. the two NaK pumps in the reflector-moderator

cooling system,

4. the four NaK pumps in the fuel cooling system,
. the two Nai pumps in the fill-and-drain tank

cooling system,
6. the four radiater blowers in the main duct,
7. the two radiator blowers in the special duct,
§. the battery for control and instrument circuits,
9. the pump lube cil systems,
10. the heater and auxiliary loads.

The relation between the various pumps is such

n

that one fuel pump, one sodium pump, one NaK
pump in the reflector-moderator cocling system,
two NaK pumps in the fuel coeling system, and
two radiator blowers should be connected to one
power source. 1[hese major pieces of equipment
will not have alternate power sources.

A station-type battery will be provided, and the
circuit will be arranged so that the battery will
float on the line at all times to keep a full charge.
The battery wiil be used to supply the necessary
control and instrument circuits in case of a power
outage. Seme emergency lighting will alsc be fed
off the bottery.

42

Fuel and Sodium Pump Lubricating
and Coocling Oil Systems

The lubricating and cooling oil systems for the
fuel and sodium pumps are to be sealed so that
they will serve as secondary, or backup, containers
for the gaseous radioactive products which might
leak from the fuel or sodium systems through the
primary rotating face seal between the reactor
system cover gas and the lubricating oil surrounding
the lower seal. The pressure in the lubricating
cil system will be a slave to the pressure of the
process system through seal-balancing control for
holding the differenticl pressure across the lower
seal to a minimum and in @ direction to cause
leckage of oil into the system rather than gases
into the oil. The oil leakage will be trapped and
then removed by a separate system. All parts of
the system external to the test cell must be
capable of withstanding pressures as high as
200 psig for periods of up te 1 hr without failure.

Process Water System

The ART process water is to be supplied by an
““open,”’ once-throcugh system with two perallel
pumps to supply pressure. There will be one supply
header and one exit header, and each will penetrate
the cell wall. The process circuits will join these
headers within the cell to minimize the number of
cell penetrations. The flow through each circuit
witl be preset before the cell is sealed, and thus
there will be no need for remote conirol or remote
flow measurement, The use of the open cycle
eliminates the need for additional water pumps,
heat exchangers, etc., and the consequent neces-
sity of canning the pumps to withstand the csll
disaster pressure. Check valves, backed up by
motor-driven cutoff valves, will be provided in the
infet lines through the cell. The check valves,
backed up by the cutoff valves, will prevent the
escape of the gaseocus activity that would be
present in the water lines within the cell if a
reactor catastrophe caused the water lines within
the cell to rupture. Provisions will be made to
assure uninterrupted flow, since water flow to the
lead shielding must be maintained during all periods
when the reactor and dump tank are at operating
temperature so that the lead wiil not be melted.

Water will be required within the cell for filling
the reactor water shieid and for ceoling the resctor
and fuel-dump-tank lead shields, the two instrument
pods, and the reactor and fuel-dump-tank vapor
sraps. Outside the cell, water cooling will be

 
 

provided for the lubrication and hydraulic systems,
the charcoal adsorber, the penthouse space cooler,
and the NalK cold treps and for filling the water
annulus of the cell.

THE FACILITY
The Building

The building constructed in 1952 te house the
ARE is being modified to provide the space and
facilities required for the ART. An addition has
been constructed at the south end of the ARE
building to effect a 64-ft extension of the eriginal
106-ft-long building, The shielded reactor assembly
will be installed in the containing cell that has
been provided in the addition for this purpose.
Such an arrangement will permit the use of services
and facilities that were provided for the original
installation., [tems such as the controf room, of-
fices, change roems, toilets, storage orea, water
supply, power supply, portions of experimental
test pits, access roads, security fencing, and
security lighting have been incorporated in ART
plans.

The plan and section drawings of the facility
are shown in Figs. 19 and 20. The floor level of
the addition is at the ARE basement-tloor grade
(ground level at this end of the building), and the
cetl for housing the reactor assembly is sunk in
the floor. The reactor cell is located in approx-
imately the southwest quarter of a 42-fi-wide by
64-ft-long high~bay extensicn and is directly in
fine with the ARE experimental bay. The reactor
assembly will be positioned so that the top of the
shield will be below building floor efevation.

The socuth wall of the ARE experimental bay has
been removed, and the overhead crane facility has
been revised by the installatien of ¢ 30-ton-capacity
crane, in addition to the existing 10-ton c¢rane, tc
permit use of the experimental pits for instailation
of auxiliory equipment and possibly for underwater
reactor disassembly work after reactor operation.
Also, the truck dosr in the north wall of the building
has been enlarged to provide a large entry door
to the ART area. Field maintenance and laboratory
facilities have been instailed in the area east of
the new bay and south of the low bay of the older
part of the building.

The Reactor Assembly Celf

The cell designed for housing the reactor as-
sembly is shown in Fig. 21. The cell consists of
an inner and an outer tank. The heat-dump equip-

ment will be located outside the cell, but nearby.
The space between the two tanks is 36 in. and
will be filled with water. The inner tank will be
sealed so that it can contain the reactor in an
inert atmeosphere of nitrogen at 5 psig. The tank
has been built to meet ASME code requirements
for unfired pressure vessels designed for a 200-
psig operating pressure. The outer tank serves
merely as a water container,

The inner tank will be approximately 24 ft in
diameter with o straight section 12 ft long and
a hemispherical bottom and top. The outer tark,
which is cylindrical, is 30 ft in diometer and about
47.5 ft in height. When the reactor is to be operated
at high power, the space between the tanks and
above the inner tank will be filled with wcter for
carrying off the heat given off by decay gamma
activity in the event of an accident so severe as
to cause a meltdown of the reactor,

About 13 ft of the outer tonk will be above fioor
grade. This portion of the tank, as well as the
top hemisphere of the inner tank, will not be
attached until completion of the reactor instcliation
and preliminary shakedown testing. Since the
shielding at the reactor and for other radicactive
components will be quite effective, it will be
possible for o man to enter the inner tank through
a marhole for inspection or repair work, |f the
reactor has been operated at moderately high power,
the fuel will have to be drained. If the repair
work requires a relatively feng time, the nitrogen
atmosphere will be replaced with air.

The unshielded reactor assembly will weigh
approximately 11,500 ib, the lead gamma shield
will weigh approximately 26,000 tb, and the water
in the shield will weigh approximately 38,500 Ib.,
The shielded reactor assembiy will be mounted in
the inner tank on vertical columns with the reactor
off center 3 ft from the vessel axis and abcut 7 ft
above an open-grated floor. This positioning pro-
vides the space that wiil be needed for the location
of the fuel fill-and-drain tank, prepower and post-
power sampling systems, fuel recovery tank, nuclear
instrumentation, and enriching equipment. The
off-center location alse serves te minimize the
length of the NaK piping.

The NaK aond off-gas piping connected to the
reactor will pass through a thimbie-type passage
or bulkheod with a bellows-type expansion joint
in the double-walled cell. The opening wili be
covered with a conical thermal sleeve which will
be welded to the pressure cell wali. The piping

43
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Fig. 19. Plan of the ART Building.

   
 

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TOR ASSEMBLY CELL

REAC

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Fig. 21. Reoctor Assembly Cell

 

46
 

 

will be anchored to the thermal sleeve., The
annular space between the NaK and the off-gas
piping end the thimble-type passage will be filled
with insulation. Auxiliaery service pipes and tubes
connected to the reactor will pass through three
junction panels located on the west side of the
cell. The openings will be covered with stiff
plates which will be welded to the pressure-cell
wall, The bulkhead between the double-walled
cell is a bellows-type seal.

Five doubly sealed junction panels for controls
and instrumentation have been installed through
the tank below the building floor grade on the east
side of the tank as a part of another bulkhead
system to pass the wires, pipes, tubes, etc,,
required for the circuits and systems, The velume
within the instrument and electrical wiring bulk-
heads will then be maintained at a pressure of
200 psig to prevent outleakage from the inner
tank, even in the event of a disaster. The various
thermocouples, power wiring, etc., will be installed
on the reactor assembly in the shop and fitted
with disconnect plugs so that they can be plugged
into the panel in a short period of time after the
reactor assembly has been lowered into position
in the test facility. This will minimize the amount
of assembly work required in the field.

Manholes 5 ft in diameter have been installed
in the upper portion of the containers. The manhole
in the water tank is located just above the flange
on the inner tank to allow passage through both

container walls and thus provide an entrance to
the inner tank for use after placement of the top.
Sufficient catwalks, ladders, and hoisting equip-
ment will be installed within the inner tank to
provide easy access for servicing all equipment.

The control tunnel surrounds 180 deg of the
north side of the cell, with all junction oanels
exiting into the control tunnel which extends to
the auxiliary equipment pit (formerly the ARE
storage pit), The pit and the adjoining basement
area will include such items as the lubricating
oil pumps and coolers, hydraulic oil pumps, relays,
switch gear, voltage regulators, and auxiliary
equipment control panels.

The Shielding Experiment Facility

Tests made at the Tower Shielding Facility
indicated that provision should be made for the
measurement of the gamma-ray spectrum of the
ART as a function of the angle of emission from
the reactor shield, Therefore five collimator tubes
have been provided to collimate four beams ra-
diating from an equatorial point at the surface of
the water shield at angles of from 0 to 70 deg from
the radial direction and one beam from an equatorial
point at the surface of the reactor pressure shell.
The latter beam will be used only during low-power
operation. The layout required for providing these
beams is shown in Fig. 22. In addition to the
facilities shown in Fig. 22, ¢ gamma-ray dosimeter
will be located on the roof above the reactor.

47
8y

   
    
     
   
    
    
  

 

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PART BASEMENT PLAN

Fig. 22. ART Shielding Experiment Facility.

 
Part [

DESIGN AND DEVELOPMENT STUDIES
 

 

DESIGN AND DEVELOPMENT STUDIES

FUEL BDEVELOPMENT

A thorough survey of materials that oppeared to
be promising as heat transfer fluids was presented
in ORNL-360.! The first requirement is that the
fluid must be liquid and thermally stable over the
temperature range from 1000 to 1800°F. A melfing
point considerably below 1000°F would be prefer-
able for ease in handling, but a substance that
would be liquid at room temperature would be even
better.  Other desirable characteristics are low
neutron absorption, low viscosity, high volumetric
specific hect, and high thermal conductivity.
Above all, it must be possible to contsin the
liquid in a good structural material at high tempera-
tures without serious corrosion or mass transfer of
the structural materiol, If the heat transfer fluid
is to serve also as a circulating fuel, it must be a
suitable vehicle for uranium; that is, the solubility
of uranium and its effect on the properties of the
fluid must be considered.

To be suitable for a high-temperature, liquid-
fueled, epithermai reactor, a fuel sysiem must meet
several stringent requirements. The liguid must
contain a sufficient concentration of a uranium
compound for a critical mass to be provided in the
core volume and must melt at o temperature sub-
stantially below the heat exchanger outlet tem-
perature. [t must consist of elements of fow ab-
sorption cross section for thermcl neutrons and
must consist solely of compounds which are
thermally stable at temperatures in excess of the
core outlet temperature. A high thermal coefficient
of expansion is desirable as an aid to selfregulation
of the power level. In addition it must be stable
to the intense radiation field ond must tolerate
the fission process and the accumulation of fission
products without serious adverse effect on its
physical, heat fransfer, and chemical properties.
Of the many fluid materials which have been con-
sidered, molten fivorides are the only fluids which
seem to be generally suitable as aircraft reactor
fuels;2 g number of fluorides of low cross-section
eiements appear to be particularly promising.?

 

A, s, Kitzes, A Discussion of Liquid Metals as
Pile Coclants, ORNL-3460 (Aug. 10, 1949).

ZW, R, Grimes and D. G. Hili, High Temperature Fuel
System, A Literature Survey, Y-657 (July 20, 1950)
W. R. Grimes et al., p 915 in The Reactor Handbook,
voi. 2, sec. 6 {1953).

A mixture of NaF, ZrF , and UF, essentially a
solution of Na,UF, in NaZrF,, containing 5.5
mole % of UF,, proved to be adequate as fuel for
the ARE. Mixtures of this general composition
are relatively noncorrosive to low-chromium nickel-
base alloys and would appear to be adequate for
use in the ART, but the melting point, vapor
pressure, and heat transfer properties are irferior
to those obtaincble from some other fluorides.

An improvement in melting point and a slight
improvement in vopor pressure and viscosity can
be obtcined by the addition of REF to the NafF-
ZrF ,-UF, system. However, the benefits obtained
would seem to be marginal, considering the high
cost of RbF and the added complexity of the
system, unless very high uranium concentrations
(above 6.5 mole %) are required for criticelity.

Substantially lower melting points and vapor
pressures con be obtained at some loss in heat
transfer properties and at o slight increase in
corrosion and moss fransfer rates with mixtures of
NaF-BeF,-UF, as fuel. Some improvemen® over
this mixture results if LiF-BeF,-UF, is used. in
view of the diminished heat transfer performance,
the toxicity of BeF,, and the cost of the required
Li7, neither mixture appears to offer any real
advantage over the NaF-ZrF -UF , system for air-
craft reactors, as presently conceived.

A slight improvement in melting poinf and a very
real advantage in vapor pressure and heat transfer
performance result from a solution of UF:4 in the
ternary eutectic of NaF-KF-LiF. Although such
mixtures are quite incompatible with Inconel and
similar commercial chromium-bearing nickel alloys
because of rapid mass transfer of chromium, the
advantages will almost certainly justify the cost of
the required Li7F if an adeguate structural and
container material can be found.

While information on the radiation behavior of
several of these general classes of fiuoride fuels
is meager, apparently there is no substantial dif-
ference among them in their stability to the radia-
tion fields and no one of them will show @ pro-
nounced advantage in response to fission or
fission-product buildup during reactor cperation.

If an adequate container material becomes avail-
able, it is very likely that the NaF-KF-LiF-UF
fuef system will be chosen for use in sircroft
reactors. As long, however, as [nconel or some
similor chromium-bearing alioy of nickel has to be

51
used as the container, the NaF-ZrFA-UF4 mixture
{or some slight variant of it) will continue to be
preferred,

The physical properties of the NaF-ZrF UF,
(50-46-4 mele %) fuet are as follows:

525°C (977°F)
1230°C (2246°F)

Melting temperature
Boiling temperature

Heat capqciiy3 {cal/g*°C)

Liquid (550 < ¢ < 850°C) 8.26 1 0.03

Solid (350 < t < 500°C) 0.2% £0.03
Heat of fusion (cal/g) 57 * 10
Thermal conéucfivify4 1.3 10,2

[Bsu/hrft 4 OF /51)]
Viscosity® {centipoises)

At 600°C 8.5

At 700°C 5.4

At 800°C 3.7

Density® {g/em?)
At 530 <z < 900°C 3.93 - $.00093:
At room temperature 4.09

Electrical v::oncdl.;r;flriw.tii*y7 {ohmecm) =1

At 1000°F 0.72
At 1200°F 1.02
At 1400°F 1.30
At 1600°F 1.56
Surface tension® {(dynes/cm)
At 530°C 158
At 600°C 139
At 733°C 115

 

Jw. D. Powers and G. C. Blalock, Enthalpies and
Heat Capacities of Solid and Molten Fluoride Mixtures,
ORML-1956 (Jan. 11, 1958).

4s. 1. Claiborne, Measurement of the Thermal Con-
ductivity of Fluoride Mixtures No. 14 and 30, ORNL
CF-53-1-232 {Jan. 8, 1953).

35. 1. Cohen and T. N. Jones, The Effect of Chemical
Purity on the Viscosity of a Molten Fluoride Mixture
ORNL CF-56-4-148 (April 17, 1956).

SThis equation should yield liquid densities that are
good te within 5% (5. I. Cohen and T. N. Jones, A
Summary of Density Measurements on Molten Fluoride
Mixtures and a Correlation Useful for Predicting Densi-
ties of Fluoride Mixtures of Known Composition, ORNL-
1702, May 14, 1954). The value measured at room tem-
perature checks with the predicted value to within 1%
{S. t. Cohen and T. N. Jones, Measwrements of the
Solid Densities of Fluoride Mixture No. 30, Ber, and
NaBeF 3, ORNL CF-53-7-126, July 23, 1953).

Accurate to within 10% (N. D. Greene, Megsurements
of the Electrical Conductivity of Molten Fluorides,

ORNL CF-54-8-64).

8Accuracy believed to be to within about £20% [S. I.
Cohen and T. N. Jenes, Preliminary Surface Tension
Measurements of the ARE Fuel (Fluoride Mixture No,

30), ORNL CF-53-3-259 {Mar. 27, 1953)].

52

STRUCTURAL MATERIAL

In selecting the structural materials for the fuel
and coolant systems (which specify the structurdl
material for the entire reactor), the key considerc-
tions were nuclear properties, high-temperature
performance, fabricability, availability, and cor-
rosion resistance fo fuel and coolant, The material
was required to have both high creep strength at
high temperatures and a ductility of at least 10%
throughout the cperating temperature range so that
high focal thermal stresses would be relieved by
plastic flow without cracking, Also, the metal
was to be highly impermeable and weidabie, with
ductility in the weld zone of ot least 10% throughout
the temperature range from the melting point to
room temperature. The availability of the material
was a prime consideration, because a good assort-
ment of bar stock, tubing, and sheet is essential in
a development program. These considerations led
to the use of Inconel for the structural material of
the ARE, ond it has been selected for the ART
because extensive research and development work
has not yet produced a superior material in sub-
stantial quantities,

The most promising of the other materials being
considered are the nickel-molybdenum alioys.
These alleys are vastly superior in corrosion re-
sistance to the fluoride mixtures and in strength
at the temperatures of interest; however, the
commercial nickel-moiybdenum alloys Hastelloys
B aond W exhibit brittleness in the temperature
range 1100 to 1500°F as a result of age-hardening.
Present indications are that this embrittlement
can be overcome by reducing the molybdenum con-
tent to 15 to 20% and conirolling the amounts of
strengthening additions, such as Al, Ti, W, Nb,
Cr, and C. [f the oge-hardening tendencies can be
eliminated without excessive compromise of
strength ond corrosion properties, structural me-
terial will become available that will be satis-
factory at operating temperatures up to 1700°F.

CORROSION OF STRUCTURAL MATERIALS

Extensive studies of the corrosion of Inconel by
molten fluorides, especially NQF-ZFF4-UF4 mi Xe
tures, have been conducted. For these studies,
static and dynamic tests were performed both out
of and in radiation fields. The tests have been
made with Inconel thermal-convection loops oper-
ated (out-of-pile} with a top temperature of 1500 to
1650°F, wall temperatures up to 1750°F, and

 
 

 

tempercture gradients of 225 to 250°F, which in-
duce flow rates of 2 to 6 fpm in.the fluoride mix-
ture. The effects of time, temperature, purity,
ratio of surface arec to volume, various additives,
valence state of the uranium, and conecentration of
the uranium have been investigated. In general,
data from the thermal-convection loops have
agreed well with the data from the forced-circulation,
high-temperature-differential loops that have been
operated out of and in radiation fields. The
velocity and Reynolds number of the circulated
fluid apparently have a very minor effect on cor-
rosion and mass transfer. |t has alsc been demon-
strated in in-pile forced-circulation loops that
radiation has little effect on these Inconel-fluoride
mixture systems at power densities up to T kw/em?,
In the most recently operated in-pile loop, the fuel
burnup reached a level essentially the same as the
design value for the ART. This loop is being
disassembled for anclysis.

inconel, a solid-solution alloy of Ni, Cr, and
Fe, when exposed to fluoride metals is subject 1o
preferential leaching of chromium, which causes
the formation of voids in the metal. The voids are
subsurface, are not interconnected, and are found
to occur both within the grains ond at grain bound-
aries, The selective leaching of the chromium
occurs not because of physical selubility of chro-
mium metal in molten fluorides but by chemical
reaction of the chromium with oxidizing agents
present in the fluoride mixture or on the original
metal surface. Accordingly, corrosion of Inconel
by the fluoride mixture is strongly dependent on
the concenfration of these reducible compounds;
this dependence is emphasized as the ratio of
fluoridesmixture volume to Inconel surface area
is increased,

Typical impurities react and produce corrosion
by the following processes:

2HF + Cr° —— GrF, + H,

NifF, + Cr®—— CrF, + Ni°
FeF2 + Cr® —> CrF, + Fe®
2FeF, + 3Cr° —— 3GrF, + 2Fe®
2CrF3 + Cr® —— 3CrF,

Oxide films on the metal walls react with the fuel
constituents (ZrF, or UF,)} fo yield structurdl
metal fluorides:

ZNi0 + ZrF4—=-——=-9- NiF, + Zréfi)2
2F6203 + 3ZeF , ——> 4FeF, + 3Zr0,
2Cr203 + 3ZrF  ——> 4CrF 4 + 3Z¢0,

These structural metal fluorides are then available
for reaction with chromium, as shown above, It is,
accordingly, necessary that the fluoride mixture
and the Inconel be of especially high purity. If
the purity specification is met and UF, is the
uranium compound used, the reaction

UF, + Cr° === CrF, + 2UF,

becomes the rate-determining reaction in the cor-
rosion process.

in the thermal-convection loops, reduction of
impurities in the fluoride mixture and equilibration
of the metal surface with the UF, seem to require
200 to 250 hr and to produce void formation to o
depth of 3 to 5 mils. In the forced-circulation
loops, in which the flow rates of the fluorids mix-
ture are much higher, these reactions proceed more
rapidly but do net cause ony greater corrosion.
For a given concentration of impurities of UIZ4 the
depth of attack varies with the ratio of surface
area to fuel volume; and, if equilibrium is estab-
lished isothermally, the corrosion is reascnably

uniform. However, in a system with a femperature
differential the hotfest zone is preferentially
attacked,

When fetravalent uranium is present, the reaction
UF, + Cr® == CrF, + 2UF,

is responsible for some ‘‘mass transfer’’ in addition
to the corrosion described cbove. In the mass
transfer process, chromium removed from the metal
walls in the hot zone is deposited on the metal
walls in the lower temperature regions of the sys-
tem. Since the equilibrium constant for the reaction
is temperoture dependent, the reaction proceeds
slightly further to the right in the high-tempercture
zone {1500°F) then in the [ow-temperature zone.
The UF; and the CrF, are soluble and therefore
move with the circulating stream to the cooler zene,
where a slight reversal of the reaction occurs and
chromium metal is formed. This type of conversion
is not readily apparent in thermal-convection loops
operated for 500 hr, because in the first 500 hr
of operation the mass fransfer effect is masked by
the effects of impurities ond nonequilibrium con-
ditions. [n loops operated for 1500 hr, however,
the moss transfer effect is observable. It is esti-
mated that, in Inconel systems circulating a zir-
conium-base fluoride fuel mixture, corrosion by
the mass transfer mechanism will increase by
about 3 mils per 1000 hr. Data from a loop oper-
ated for over 8000 hr substantiate this view; the
total depth of attack was 25 mils, with a peak
Inconel temperature of 1500°F,

33
The corrosion attack in the ART will be greatest
in the hottest portion of the fuel circuit, and aofter
1000 hr of operation the attack is expected to be
about 8 to 10 mils if all the uremium is present
originally as UF .. If a mixture of UF, and UF,
is used, the corrosion of Inconel by the fue! will
be decreased., For instance, after 500 hr of circu-
lation of @ UF ;-containing fused salt, attack of 1
to 2 mils is found, in contrast to the usual 3 to
5 mils. Experimenis are under way for defermining
the proper UFB«*M«UF4 ratic for achieving minimum
cotrosion along with adequate solubility in the
zirconium-base fluoride mixture. The UF. is not
sufficiently soluble in the NoF-ZrF, carrier to
provide a critical mass, ond plating-out of the
uronium weuld also be a problems

it has been established that the zirconium-base
flucride mixture fuels and Inconel will be com-
patible under ART operating conditions for the
1000-hr expected life and that the ottack will not
seriously weaken the reactor structure, but there
is still some concern about the corrosion of the
thin-walled (0.025-in.-thick} heat exchanger tubing.
The cmount of mass transfer of chromium to the
cold leg will be so small that there will not be an
increase in pressure drop or ¢ decrease in heat
transfer performance.

The materials compatibility problem has also
been investigoted for the materials of the reflector-
moderator system in which sodium will flow in
direct contact with both berytlium and Inconel.
fn such @ system both temperature gradient mass
transferand dissimilar metal mass transfer between
the beryilium and Inconel can oceur. On the basis
of numercus whirligig, thermal-convection loop,
and forced-circulation loop tests conducted on
beryllium-sedium-inconel systems, the former does
not appear to be a serious problem if the tempera-
ture is kept below 1300°F. The temperature
gradient mass transfer detected on the cold-leg
walls of any thermal-convection loop with a beryl-
lium insert in the hot leg operated at 1300°F (cold
leg, 1100°F) for periods of 1000 hr has been less
than 200 ug per square centimeter of beryilium,
The mass fransfer of Incenel by the sedium at
temperctures below 1300°F is not considered to be
a serious problem, A beryllium-sodium-Inconel
forced-circulation loop in which egual creas of
inconel and beryllium were exposed to the sodium
showed no incresse in mass fronsfer over that
found with ali-Inconel loops. This loop was oper-
ated ot ¢ hot-leg temperature of 1300°F for 1000 hr,

54

and a 2-mi! deposit of crystals was found in the
coid leg. The crystals were nickels=rich and con-
tained little or no beryllium,

Dissimilor metal mass transfer, in which the
socdium ccts as a carrier between the beryllium
and Inconel, hos been found to be o function of the
separation distance between the two materials.
When the beryllium and Inconel are separated by
distances of greater than 5 mils, no continuous
layers of nickel-bery{lium compound are formed on
the surface of the Inconel in 1000 hr at 1200°F.
Under these conditions a fine subsurface pre-
cipitate, probably BeNi, forms to a depth of approx-
imately 2 mils with ¢ 5-mil separation between the
beryltium and the Inconel and to a depth of 1 mil
with a 50-mil separation.

The major compatibility problem in the reflector-
moderator system will occur in those areas where
inconel and beryllium are in direct contact. In
1000 hr ot 1200°F o 5-mil layer of brittle compound
(4.5 mils of Be, Ni.; 0.5 mil of BeNi} formed on
the Inconel when it was held in direct contact
with beryllium with no applied pressure. Approxi-
mately 1 mil of Inconel was consumed in the forma-
tion of this layer. At 1300°F under a contact
pressure of 500 psi a 25-mil layer of Be, Ni, +
BeNi, which consumed 4 to 5 mils of Inconel, was
formed on the Inconel in 1000 hr. One pessible
sclution te this problem may be to chromium-plate
the Inconel. Tests under the above conditions
{1300°F, 500 psi, 1000 hr) indicate that a S-mil
chromium plate can reduce the reaction to a cone-
sumption of 2 mils of Inconel.

Since the siructural metal of the fuel system will
also be in contact with the sodium in the NaK in
the heat exchange systems, intensive studies are
under way on the competibility of inconel and
sodium under dynamic conditions at ¢ peak sodium
temperature of 1500°F. It is believed that mass
transfer in these systems can be kept fo a tolerable
minimum by the use of very pure, non-oxygen-
containing sodium and very clean metal surfaces.
In  particular, it is expected that the beryllium
surfaces will remove the oxygen impurity from the
sodium and thus contribute to the reduction of mass
transfer in the sedium system.

Other structural materials have been studied or
are being studied in an attempt to find o material
superior to Inconel for subsequent aircraft reactors,
The stainiess steels appear to be superior in
sodium, but they are definitely inferior in the
fluoride mixtures. Hastfelloy B is far superior in

 
 

 

the fluoride mixtures, but it has poor fabricetional
qualities.  Various modifications of the basic
nickel-molybdenum alloys are now being investi-
goted.

RADIATION EFFECTS ON STRUCTURAL
MATERIALS

The effects of irradiation on the corrosion of
Inconel exposed to a fluoride fuel mixture and on
the physical and chemical stability of the fuel
mixture hove been investigated by irradiating
Incenel capsules filled with static fuel in the MTR
and by operating in-pile forced-circulation Inconel
loops in the LITR and in the MTR. The relatively
simple capsule tests have been used extensively
for the evaluation of new matericls. The principal
variables in these tests have been flux, fission
power, time, and temperoture. In a fixed neutron
flux the fission power was varied by adjusting the
uranium content of the fuel mixture. Thermal-
neutron fluxes ranging from 1017 to 104 neu-
tronsscm=2esec—! and fission-power levels of
80 tc 8000 w/cm?® have been used in these tests.
Almost ail the capsules have been irradiated for
300 hr, but in some of the recent tests the irradia-
tion period was 600 to 800 hr. After irradiation the
effects on the fuel mixture were studied by meas-
uring the pressure of the evolved gas, by determin-
ing the melting point of the fuel mixture, and by
making petrographic and chemical analyses. The
Inconel capsule was also examined for corrosion
by standard metallographic techniques.

In the many capsuie tests made to date no major
changes that can be attributed to irradiatior, other
than the normal burnup of the uranium, have occurred
in the fuel mixtures. However, the anclytical
method for the determinction of chromium in the
irradiated fuel mixture is being rechecked for
accuracy. The metaliographic examinations of
Inconel capsules tested at 1500°F for 300 hr have
shown the corrosion to be comparable to the corro-
sion found under similar conditions in unirradiated
capsules, that is, penetration to a depth of less
than 4 mils. In capsules tested at a temperature
of 2000°F and above, the penetration was tc a
depth of more than 12 mils and there was grein
growth.

Three types of forced-circulation in-pile locps
have been tested. A large loop was opercted in a
horizontal beam hole of the LITR. The pump for
circulating the fuel in this loop was placed outside
the reactor shield. A smaller loop, including the
pump, was operated in a verfical positien in the
lattice of the LITR. A third loop was oserated
completely within a beam hole of the MTK., The
operating conditions for these loops are presented
in Table 6, and results of chemical onalyses of the
fuel mixtures circulated are given in Table 7.

The LITR horizontal loop operated for 545 hr,
cncluding 475 hr ot full reactor power, The loep
generated 2.8 kw, with a maximum fission power of
400 w/em®, The Reynolds number of the circulated
fuel was 5000, and there was o temperature differ»
ential in the fuel system of 30°F, The voiume of

TABLE 6. OPERATING CONDITIONS FOR {HCONEL FORCED-CIRCULATION IN-PILE LOOPS

 

 

Operating Varicbles LITR Horizontel LITR Vertical MTR In-Pile
Loop Loop l.osp No. 3

Na FoZrFA-U F, composition, mole % 62.512,5-25 63-25-12 53.50406.5
Maximum fission power, w/cm’ 400 500 730
Total power, kw 2.8 5.0 29
Dilution factor 180 10 3.5
Maximum fuel temperature, °F 1500 1500 1500
Temperature differenticl, °F 30 71 175-200
Reynclds number of fuel 5000 3000 5000
Operating time, hr 645 130 462
Time at full power, hr 475 30 271
Depth of corrosion attack, mil 1 i 1

 

55
TABLE 7. CHEMICAL ANALYSES OF FUEL MIXTURES CIRCULATED IN INCONEL
FORCED-CIRCULATION IN-PILE LOGPS

 

Loop Designation Sample Taken

Minor Constituents (ppm)

 

 

fron Chromium Nickel
LITR horizontal loop Before filling 80 + 10 10t5 200 £ 100
After draining 180 T 40 150 + 10 0ts
LITR vertical loop Before filling 20 10 80+ 10 145+ 20
After draining 37¢ 1 20 100 *+ 2¢ 50 £10
MTR inepile loop No. 3 Before filling 40 £ 10 60 £ 10 40 + 10
After draining 240 + 20 50 £ 10 100+ 20

 

the loop was large, and therefore there was a
large dilution facter, that is, the ratio of the total
system volume to the volume in the fissioning
zone. Metaliographic analyses showed that there
was less than 1 mil of corrosion of the Inconel
walls of the loop. Chemical analyses showed that
the irradiated fuel mixture contained 200 ppm or
less Fe, Cr, and Ni. Therefore there was ne evi-
dence of accelerated corrosion in this experiment,

The LITR vertical loop cperated for 130 hr, with
only 30 hr of the total operating period being at
full reactor power. The loop generated 5 kw, with
a maximum fission power of 500 w/em3. The
Reynolds number of the fuel was 3000, and the
temperature differential was 71°F., The surface-
to-volume ratic was 20, and the dilution factor was
10. Chemical analyses showed that the irradiated
fuel contained 370 ppm Fe, 100 ppm Cr, and 50 ppm
Ni, Metallographic analysis of the loep showed
that there was less than 1 mil of corrosion at the
curved tip.

The horizontal loop inserted in the MTR (MTR
in-piie loop No. 3)7 operated for 462 hr, including
271 hr at power. The locp generated 29 kw, with
a maximum power of 730 w/ecm3. The Reynoids
number of the fuel was 5000, ond the temperature
differential was 175 to 200°F for 103 hr and was
T00°F for 168 hr, The dilution factor was about
3.5.

Chemical analyses of fuel from MTR in-pile loop
No. 3 showed that it contained 240 ppm Fe, 50 ppm
Cr, and 100 ppm Ni. The high iron concentration
wos prebably caused by o sampling difficulty.
Examingtions of unetched metallographic sections

 

°D. B. Trauger et al, ANP Quar. Prog. Rep. Dec. 10,
1955, ORNL-2012, p 27.

36

of the lnconel tubing showed that there was no
corrosion penefration; etching revealed no attack
tc a depth of more thon 1 mil. A slight amount of
intergranular void formation was noted hut was
neither dense nor deep. Measurements of wall
thickness did not reveal any variations attributable
to corrosion. The loop was examined carefully
for effects of temperature variations between the
inside ond outside walls of tubing at bends, but
no effects of overheating were observed. Samples
taken from the inlet, the center, and the exit side
of the loop had less than 1 mil of corrosion. The
low corresien is credited to the careful temperature
contrel of the salt-metal interface and to the maxi~
mum wall temperature being below 1500°F at ail
times,

Future loops will be cperated at higher fission
powers and therefore greater temperature differ-
entials. The dilution factors will be kept low.
New fuels and new alloys are being censidered
for testing in future loops.

STRUCTURAL DESIGN ANALYSES
The Reactor

The five principal probiems involved in the
structural design analysis of the ART are the
selection of the operating conditions for the reac-
tor and the definition of the accidents or failures
that might occur for which corrective action would
be possible; the determination of the mechanical
or pressure loads set up by these various operat-
ing conditions; the calculation of the correspond-
ing temperature distributions throughout the reac-
tor; the selection of suitable design criteria; and
the detailed stress analysis and structural design
of the system,

 
:{1:3;-;-:-:-:‘-,:"‘

The proposed operating program for the ART is
based en the reactor being at temperature {(1200°F
or greater) for 1500 hr. During the last 1100 hr of
this time, the recctor will be critical and will be
subjected to 25 full-power cycles to simulate
flight requirements. In general, the most severe
steady-state pressure loads to which the internal
structure of the reactor will be exposed will occur
during full-power operation, '% which is referred to

' and which serves

as the "‘design-pcint condition
as the basis for steady-state-load design analysis.
Design for transient loads is bused in large part
on the power-cycling conditions mentioned above
and on various unscheduled changes in power level
that will occur as corrective or safeguard action in
the event of certain accidents. Not all the pos-
sible operational situations which may arise during
the test of the ART have yet been examined from a
design viewpoint, but it is believed that those
situations which will impose the most severe re-
quirements on the structural design have already
been considered. However, if further studies
indicate the contrary, modifications will be made
in the design or constraints will be imposed on
the operational procedure. Information on tem-
percture and pressure transients to be expected
from off-design operation will be obtained frem
tests performed on the ART simulator {(e.g., abrupt
stoppage of several pumps).

The pressure loads within the reactor have been
determined for the design-point condition and for a
condition involving the failure of ene fuel pump.
It is believed that these two situations represent
the most severe symmetric and unsymmetric in-
ternal leads to which the reactor will be exposed.
The calculation of the one-pump-out condition was
performed in order to determine the stability of the
reflector-moderator assembly. Under this circum-
stance there would be no presswe drop through
the fuel-to-NaK heat exchangers in the circuit
supplied by the pump thet failed. This conditien
would result, then, in a net side load on the re-
flector of 12,400 ib. The reflector support ring
has been designed to be stable against the re-
sultant overturning moment, and the assembiy can
be expected te remain seated against the north-
head structure.

 

0The main heat exchanger headers and the core
shells ore two important exceptions. The most severe
loads on these members will occur in the event of the
stoppage of a fuel pump.

Feor design-point operation the pressures in the
fuel and in the sodium circuits will be symmetric
with respect to the vertical axis of the reactor;
thus all resyltant forces on the principal struc-
tural components (the reflector, north head, island,
pressure shell, and pressure-shell liner) will be
directed along this axis. In oddition to these
vertical forces, horizontal reactions will ocecur
between the reflector shell and the pressure-sheil
liner and between the heot exchanger tubes and
the reflector shells and the pressure-shell liner as
a result of pressure differences and differential
thermal expansion. In the heat exchanger tubes
and the thin core and reflector shells these forces
can cause large deflections and possibly buckling.

The principal forces which must be accommo-
dated at the design-point condition are the vertical
loads thot will be imposed by the fuel. These
loads consist of three components: forces result-
ing from pressure drops in the fuel passages,
weight forces, and buoyant forces. In the reflec-
tor-moderator assembly the pressure-drop force
will be 58,000 ib, the weight 3,000 Ib, und the
buoyant force 4,700 [b. The net force wili be up-
ward and will press the reflector against the north
head, which, in turn, will transmit the force to the
pressure shell,

There will also be an upward thrust of 20,900 Ib
on the north-head region of the pressure shell from
the combined action of pressure, weight, and
buoyant forces on the heat exchanger. The heat
exchanger and reflector forces will be brought into
balance through the action of corresponding pres-
sure forces on the inside of the pressure shell
(transmitted by the liner), which will force the
pressure shell downward (Fig. 23). Similar buoy-
ant, weight, ond pressure-drop forces will alseo act
on the island assembly, but these forces will be
relatively small and can be carried by the shell
structure without the aid of special structural
members.

Caleulations are being made for determining the
three-dimensional temperature distributions through-
out the sodium circuit, the fuel, and the principal
structural members for the various operating condi-
tions of the reactor. In the initial design studies,
thermal loads and stress values were used that
were based on very preliminary estimates of the
temperature structure, and therefore the very de-
tailed analysis being made will serve as ¢ check
on the ceolant-flow provisions: |t will also pro-
vide more accurate estimates of the thermal-stress
distribution in critical members.

57
85

 

NORTH HEAD

 

7/ [y \
o7 TN
i {"'fz J/ [ \'L VoA “\;"\ ““\
! : 3 |
b “.\ P ! ;" ! JH‘ f,
By \\; vy / r/ { i}.-

N
N

. UPTHRUST FROM REFLECTOR

(59,656 Ib)

. FUFL PUMP DISCHARGE PRESSURE

FORCE {340¢€ Ib}

. FUEL PUMP SUCTION PRESSURE

FORCE (7747 Ib)

. SODIUM PRESSURE FORCE ON

NORTH HEAD {psi)

. NORTH-HEAD WET WEIGHT {1000 Ib)
. PRESSURE SHELL REACTION

LLINER

 

G

=T X & H

. HEAT EXCHANGER PRESSURE DROP

FORCE (24,216 ib)

. FUEL PRESSURE FORCE FROM CORE

EXIT (65,264 b}

. STATIC HEAD (B343 \b}

. SODIUM PRESSURE FORCE

. LINER WEIGHT (1000 b)

. ISLAND (200 ib)

. PRESSURE SHELL SUPPORTING

REACTOR {J+98,993 ib)

Fig. 23. Pressure L.oad Distributions ot Full Power.

»w o O T o

 

GSESRET
ORN|-LR—DWG 16018

PRESSURE SHELL

. NORTH-HEAD RZACTION {68,809 ib)

. LINER LOADS (M)}

. MEAT EXCHANGER LOAD (20,912 Ib}

. PRESSURE SHELL WEIGHT (3500 ib)

. PUMPS (4200 Ib)

. SUPPORT LOAD ON REACTOR {(REACTOR

WEIGHT = 44,000 ib)

 
 

 

Precise calculations of the temperature profiles
throughout the reactor require detailed information
on the energy deposition from radiation and neu-
tron reactions and on the temperature structure
along the boundaries of the fiuid passages. In-
formation on the energy depesition is to be ob-
tained by two independent methods: a buildup-
factor technique in preparation at Pratt & Whitney
and a Monte Carle method developed at ORNL.
The nuclear reactien dats required for these pro-
grams are based on the two-dimensional multigroup
flux calculations suvpplied by the Curtiss-Wright
Corporation. The dual program has been initiated
because the Monte Carlo method, althcugh more
accurate, will be several months in preparation,
and it is believed that the relatively quick Prott &
Whitney method will suffice for preliminary three-
dimensional data. Finally, @ more accurate picture
of the temperature profiles through the fuel will
be obtained from analyses of the results of the
high-temperature critical experiment and of the
tests of the half-scale model of the ART core,
which wutilizes a velume heat source to simulote
the fission heating.

The utilization of the tempercture and load dis-
tribution information in the design cnalysis of the
reactor system poses some difficulties in a broad
technological sense because of the lack of an
established design philosophy for high-temperature
operation. In general, relatively little design ex-
perience has been acquired in providing for the
effects of thermal cycling, strain cycling, creep
buckling, and thermal relaxation.

Suitable design criteria are being formulated for
the ART by combining informotion obtained from

mafterials testing programs, component testing
programs, and the Engineering Test Unit (see
Part 1V}). In the materials testing programs, data

are being obtained on the properties of Inconel
and beryllium at the temperatures of interest in
order to acqguire insight intc the nature of the
basic thermal phenomenon involved.

Data on the creep and tensile properties of the
two materials, their behavior under strain cycling,
and their relexation properties are of particular
interest. The creep and strength data!! are re-
quired for the design of members that will be sub-
jected to continucus loads over prolonged periods
_of time {e.g., loads resulting from operating pres-

 

ViThis work is under way at ORML and The Brush
Beryllium Co.

The strain-cycling data'? are being cor-
13 and

sures).
related according to the Coffin formulation
will serve as the basis for the design analysis of
structures which will be repeatedly subjected to
mechanical or thermal loads that will produce
plastic deformation in the material. The reloxation
data'4 are to be used mainly to determine the
amourt of plastic strain developed under cyclic
loads. The creep-buckling information!® is re-
quired for the design analysis of the various core
and reflector shells. These shells will be sub-
jected to pressure differentials and to temperature
gradienits both through their thickness cnd along
their surfaces.

In some instances the matericals test information
obtained from the programs mentioned above may
be used directly to estimote the thermal deforma-
tion and buckling characteristics of components
and to determine the life of parts subjected to
cyclic loads. However, this direct application wili
be effective only for relatively simple struciural
configurations and cannot be used for many of the
important structural members of the reactor. In
the latter cases it will be necessary to resert to
component or scale-mode! tests under operating
conditions. For example, a test is under way on ¢
one-fourth-scale model of the cuter core shell to
determine whether the core shells will survive the
thermal cycling to which they will be exposed
during the operational life of the reactor. The
eperational conditiens for this test were based on
the anticipated program of the ART, and the tem-
peratures and hold-times involved were defermined
from relaxation and strain-cycling data on Inconel
(Fig. 24). This test should indicate, also, the
extent to which the strain-cycling and relaxation
data based on simple uniaxicl stress conditions
can be extropolated to the more complex patierns
encountered in the actual design.

The detailed stress analysis of the ART, now
under way, consists in examining the system op-
erating pressures and thermal and cyclic loads.
The operating pressure loads were used to size
the principal structural members. The criterion

 

1297his work is under way at ORNL and at the Uni-
versity of Alobama,

13, F. Cofttin, Trans, Am. Soc. Mech. Engrs. 76,
931950 (1954).

"“To be supplied by ORNL, WADC, and the Uni-
versity of Michigan.

15Bata now being collected at Pratt & Whitney and the
University of Syracuse.

59
 

 
 
   

 

SECRET
ORNL-LR—DWG 16049

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

25,000 e ——— — it —
- TOTAL STRAN 0.3 % (og = 54,000 ps) :
;
20,000 - - R e L e
F—TOTAL STRAIN 0.3 % (o, = 54,000 psi} | | |
: . ‘ i
15,000 i _ e el e e h_.+___. N ;
B FULL POWER
i\ e HOLD |
I« L : !
10,000 M R . o e e
| _~TEST CYCLE -~ ART CORE SHELL
[~ z L
- 5,000 \\ i,‘:I"ffgo TQ 60 Mw N - i’M e
o sy '
a he- HEAT-UP
w O bF——""0- 0 TO 10 Mw —— - —-- - —_—b
- ; ! ‘ | - COOL DOWN
5,000 |—— #‘"",Q;QO&,QQWN o _ o
! -—‘-_—-_4
% . //
- |
40,000 |-~ -+ i Lo T : ——— - A~
1T ISOTHERMAL _ | [1SOTHERMAL
o o
-15,000 | _A+JIF S : . R L e b 1 --H.
-20,000 b—— ————— e e | e e e
|
-25000 — - --- e — gl Lo
0 1 2 3 4 5 6 7 8 9 10 1 16 17 18 19
TIME {hr)

Fig. 24. Conditions for Thermal Cycling Tests of One-Fourth-Scale Model of ART Core Shell in

Comparison with ART Thermal-Cyecling Conditions.

presently in use requires that the stress level
created by these loads be no more than one-fourth
the value known to cause rupture from creep if
applied for 1500 hr. The stress analysis is based
on elastic theory, but for very complicated con-
figurations the analysis is supported by experi-
mental studies on modeis of the actual member.
This scheme of analysis has been applied in a
preliminary manner to all structural components of
the reactor, its support structure, the NaK piping
inside and outside the pressure cell, and the heat
dump systems {NoK pumps and radiators). These
preliminary studies have been completed, and
attention is now being focused con the detailed
stress analysis of the major structural members
which are a part of the reactor proper; these mem-
bers include the north-head double-deck composite
structure, the reflector suppert ring, the core and
reflector shells, ond the fuel dump tank and
supports.

A tentative set of design criteria has been se-
lected for the detailed examination of the system
from the viewpoint of thermal and cyelic loads.
For members that will be subjected to thermal
foads which may occur only a few times during
their life, the design criterion is that the strains
produced by these loads not exceed 0.2%. Based
on the concept of an elastic stress-strain relation,
this corresponds to a stress of around 30,000 psi
at 1300°F. Such a criterion, although in agreement
with the design philosophy of the ARE, is believed
to be extremely conservative, and o more realistic
one is being considered and will be evaluated by
the results obtained from the strain-cycling fests
for the lower number of cycles. The criterion for
the design of members subjected to cyclic plastic
strain is also based on the requirement that the
maximum deformation during a cycle not exceed
0.2%. At temperctures of the order of 1300°F «

specimen thus loaded should survive more than

 
 

 

200 cycles. This figure is based on the best
strain-cycling data now available. '  This crite-
rion will be used until more reliable information is
obtained from the strain-cycling test program under
way at ORNL ond ot the University of Alabama.
The thermal-stress criteria outlined here are in
use in the analysis of the cyelic heating of the
core and reflector shells, the pressure-shell liner,
and the thermal sleeve attachments for all pipe
cutlets from the reactor, as well as in the thermal-
stress analysis of the fuel pumps and of the NaK
manifolding ot the reactor. An anclysis is alse in
progress for determining the thermal expansions
and distortions of the various shells and the
beryllium within the reactor. These calculations
require the temperature disiributions previoustly
menticned, and the results obfained from the work
will determine the final tolerances and clearances
between these members. It should be recognized
that the proper fit of the parts under the operating
conditions will define the cold dimensions to
which the system must be assembled.

The cnalytical studies described above are sup-
ported in many creas by paraliel programs of ex-
perimental stress analysis. The bulk of the work
is being carried out at the University of Tennessee
and inciudes the octual model tests of the north-
head composite-deck sfructure, the pump barrel
and NcK pipe aftachments to the pressure shell,
the main and auxiliery heat exchanger headers, the
reflecior support ring, the blowout patch for the
pressure shell, and the NaK piping systems out-
side the reactor cell.

The emphasis in the design analysis under way
and that programed for the near future is placed on
the re-evaluation of the design criteria, the de-
tailed stress analysis of the primary structure, the
completion of strain-cycling tests of component
models, and the study of off-design and transient
operating conditions.

Auxiliary Components

Several major structural compenents external fo
the reactor, including the reactor support system,
the fuel fill-and-drain tank and support, and the
main NaK piping which carries the heat from the
reactor to the radiators, have also been analyzed.

 

Y nuclear Propulsion Program Engineering Progress
Report No. 18, October 1, 1955—-December 31, 1955,
PWAC-554.

The reactor is to be suspended from an overhead
bridgelike structure by means of the four pump
barrels {(Fig. 25). The attachment of the individual
barrels to the bridge aliows herizontal motion of
the barrels so as to ccecommodate the relative
thermal growth between the reactor {at operating
temperature) and the bridge (at room temperature).
Vertical motion of the barrels is completely re-
strained. The bridge is fixed at each end fo ©
flexible column consisting of Il-in.-thick steel
plates, 28 in. wide and 123 in. long. The [ocd
carried by each column is approximately 45,000 1b.
This value is somewhere between one-fourth and
one-half the velue required to cripple the column,

The principol function of the flexible celumns
is to aliow complete freedom for the NalC lines
to expand in going from room temperature to the
operating temperciure of the reactor. During full-
power operation the upper row of NaK lines will
be at 1070°F, the lower row at 1500°F. This will
result in o horizontal growth of 3{4 in. in the upper
fines and '55’8 in. in the lewer lines. [f the reactor
is mounted in the cold condition precisely cver the
center line of the column bases, these expansions
will translate and rotate the reactor out of the
neutral position and thereby introduce bending
stresses into the columns (Fig. 26). In crder to
eliminate the bending stresses, it is planned te
precut the NaK lines so that at room temperature
the reactor wiil be lecated 1 in. off the neutral
position toward the cell wall through which the
NaK lines enter. As the reactor heats up, these
fines will expand end move the reactor into the
neutral position, thus removing the bending loads
on the columns.

Although the flexible columns can allow for the
gress expansions of the NaK lines, they cannot
provide for the differential expansion between the
fines of any one row. In order te provide some
margin for operational accidents and freedom in
controlling NaK temperatures, it is plenned to add
several bends into each line to accommadate 300
to 400°F temperature differences between adjacent
lines. The piping layout proposed for the reactor
celt for this purpose is shown in Fig. 27.

The basic design features of the NaK piping out-
side the reactor cell are similar to those inside
the cell. The principal design reguirement is that
the pipe supports and end aftachments have suf-
ficient flexibility for the lines to expand in going
from room temperature to the operating temperature
without being subjected to excessive thermal

61
 

 

 

 

 

 

 

 

 

o .
| -—m}, Lrw v

|
-

-

|
S

*Wil ER SHIELD CONTAINER
\\
" FLEXIBLE COLUMN

 

 

 

" | REACTOR \ ‘
,r,u.__r__._ ][ _I;,

- 5 ft 8% in—— -

[
i
1
|
!
/
|
I
!
!
.f

=t

FONELCENTA
ORKL~LR—~DWG 16123

 

 

 

 

 

    
 

 

 

 

 

 

 

 

 

 

i ////

AT TEMPERATURE, WITHOUT
PRECUTTING OF NoK LINES

(o)

62

 

6|n ‘
—.18in.
S S
6|n E
—9ft Oin- - e
101 ;3 in. REACTOR CELL]
-‘:_-—_—‘"‘“_:1:%— _____ ¢
S— NoK LINES 1
121t 8'9 in.
5” olyin
Lt o _
Fig. 25. ART Support Structure.
SONPITENTIAY.
ORNL-LR-DWG 16124

_~BRIDGE

 

~~PUMP BARRELS

- REACTOR

"~

TN COLUMNS

 

 

 

D

 

AT TEMPERATURE,WITH
WITH NaK LINES PRECUT

PRECUTTING OF N¢K LINES
(&)

(c}

Fig. 26. Movement of Reactor Due 1o Expansion of NaK Lines.

POSITION OF REACTQR AT ROOM TEMPERATURE,

 
 

 

SECEELL’
ORNL-LR-DWG #:022

CELL WALL ~__

|
|

o = o

Fig. 27. Layout of NaK Piping in Reactor Cell.

stresses and without producing excessive loads on
the radiators and pump casings. One of the NaK
radiator-pump assemblies in the ART system is
shown in Fig. 28. The individual lines are welded
at the reactor cell wall, and the joints may be
considered as fixed points in the piping circuit.
The radiator ends of the lines are welded to their
respective radiators, and the entire pump-radiator
assembly is suspended by a system of spring
hangers from overhead sc as to allow freedom of

motion in the principal direction of the NaK lines.
This freedom allows for the over-all growth of the
lines, and the various bends in the individual lines
provide increased flexibility to accommodate dif-
ferences in line lengths.

The operation and flew characteristics of the
fuel fill-ond-drain tank system are discussed in
Part II. Since this tank is to serve as an eversafe
depository for the fuel, it must survive some
2000 hr of operation at temperature and possibly

63
Eanrronmrat
ORML- _R-D¥G 13202

 
 
  

gt sk in.

      

e PR |
[ i TN

e e e
= e
f Y

 

 

 

 

 

 

Fig. 28. NoK-to-Air Radiator, NaK Pump, and
NaK Piping External to Cell.

several fast dumps. The principal structural re-
quirements are determined by the 40-psi NaK pump-
ing pressure, which will produce creep in the tank
structure. The design criteria, then, are based on

creep-rupture and thermal-shock censiderations.

The most sensitive areas of the design are the
joints between the tube header sheets (Fig. 16)
and the inner cylinder. The discontinuity stresses
in these regions are of the order of 6000 psi, but
the stresses are expected to decay rapidly, once
the system comes to temperature, as a result of

64

- shell.

the stress-relaxation phenomenon. Since this com-
ponient is of vital importance to the over-all safety
of the experiment, it is planned to test the entire
tank system by using one of the NaK circuits,
This test will determine the adequacy of the de-
sign in withstanding the creep and thermal shock
effects mentioned cbove.

The support of the dump tank assembly is ac-
complished with the «id of a nitrogen cylinder
located beneath the tank (Fig. 29). Although the
tank is oftached rigidly to the reactor pressure
shell through the fuel drain line, the major portion
of the tank weight is not ailowed to bear on the
shell. The total weight of the tank, including the
fuel, is approximately 6000 Ib, of which 5000 Ib
This
arrangement introduces some complication in re-
gard to the stability of the support system, but an
analysis has shown that the proposed arrangement

will be carried by the nitrogen cylinder.

of lever arms, the structural stiffness of the drain
fines, and the weights are well within the stability
limits of the system.

RADIATICON HEATING ON THE ART
EQUATORIAL PLANE IN THE VICINITY OF
THE FUEL-TO-NaK HEAT EXCHANGER

The radiation heating to be expected in the ART
was calculated so as to provide a basis for the
design of cooling systems. The results of the
calculations of the radiation heating on the ART
equatorial plane in the outer 3 cm of the beryllium
reflector and in the Inconel and the boron-contain-
ing shells on both sides of the fuel-to-NaK heat
exchanger are presented in Figs. 30 and 31. The
total gamma-ray heating in each region is given in
Fig. 30, as well as the heating from the sources
which are the main contributors to the total in each

The encircled numbers on Fig. 30 refer to
the sources described in Table 8.

The data on heating in the copper-boron layer by
alpha particles from the B1%(n,a)Li? reaction are
plotted in Fig. 31. The heating goes to infinity at
the face of the layer closest to the core because
the heating at various points is governed by an

E | function,
® dA
o | e 2
1 A

where A is the mean free path. The integral under
the curve will be finite.

 
 

SOMNFHDENTTAL™
ORNL—-LR-DWG 16127

    
    
   

SUPPORT BRiIDGE

 

 

 

P
\\\\y —— ] - ’/ %
|: b,
il d
\\ .

- DUMP VALVE

VALVE ACTUATOR ———— 5.,

 

 

 

 

(NGK SUPPLY ?

W

- (/R

ATTACHMENT - mome——=-=
TO FLOOR

Fig. 29. Fuel Fill-and-Drein Tank Support.

65
99

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

SRERE L
. OR/NL-LR- DG 14916
10 ———y
i
- STAINLESS STEFL
g
9
TOTAL FEATING.
——{0—— HEATING FROM SCURCE . {SEE
FiUFL GARP - 1 eal, TAELE 1.2 1.
@ HEATING FROM SOURCES 1,2, AND |
" A ;i . .
i & 0 - 3 (SEE TABLE 4.2.4). |
; 2 = _
! & < % = ALl DIMENSIONS IN cm., .
: L w L2 o -
; ] o = T —l @
i L b - o L o
; Lr;;‘ = d = E - = Pt
i [¥a = 5 S |
7o = o 8. & 1. \ W 3 2 .
! % = ol O L = = =
: =’ o g , - < = O
: _ o 2 m & o x i
; 1 N = O i HETL A :
; o it = A HEL UM :
‘ > « x m ” == 30DIUY i
I o GAP P i
il 0 o o
- m i ~ -
- O ] P oA I
T8 < -« o 457 T i
8 - 057 = 34 ;
s o e o - 0457 = 0.318~ \ 5 -
= - z ] e ol baCmd G ™ bl O P A e D
= 2 S % Q S ] o= D] e ] 2.4 —
T‘Z < i ,,; o L 3 & .
= = — i
T 5 1 i i
O @ L
o =
= S
E 3
1
= o
S 4 — - T R
F; o3 o
o i T
z w
oy Ll
i 7 o
o l @
3 L _ - 4 N b , .8
| ¢ )
: 1
| 1) L
: STAINLESS STEEL 4~ ™1 \l’F‘. g
i \‘:'A\ =
=
o I . ——- 41 . o 4\’—5‘ o
g/ . e —— e - - _t:& . .
o
o~ =~
' N AV G~
. — < \ o Gd -
® 2 (G
2
@\ N A8 //@ ®\ 55 v
\ : e e~
- (G;

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

Fig. 30. Gamma-Ray Heating in the Vicinity of the Fuel-to-NaK Heat Exchanger on the E quatorial Plaone of the ART.

   
 

TABLE 8. SOURCES OF RADIATION HEATING CONSIDERED IN CALCULATING THE
RESUL TS PRESENTED IN FIG. 30

 

 

Source Source
No Source Strengih
1 Prompt gamma rays in the fuel region of the core of the reactor 28.3 w/cm3
2 Decay gamma rays in the fuel region of the core of the reactor 6.84 w/cm3
3 Gamma rays from inelastic scattering of neutrons in the fuel region of the core 10.1 w/cm3
4 Capture gammc rays in the cuter core shell 41.4 w/t:m2
5 Capture gamma rays in the reflector (average) 0.5 w/em®
6 Capture gamma rays in the first Incone! shell cutside the beryllium reflector 22.5 W/cm3
7 Boron capture gamma rays in copper-boron layer 1.8 W/t:m2
8 Alpho particles from the Bm(n,a)Lij reaction in the copper-boron layer (average) 42 w/cm:?'
g Decay gamma radiation from the fuel in the heat exchanger 2.3 w/’c:m3
10 Gamma rays from inelastic scattering of neutrons in first 9 em of reflector 0.7 w/«:.m3
{avercge)
11 Capture gamma rays from delayed neutrons in the heat exchanger and Inconel C.1 w/t:'rn3
shells {including the pressure sheil)
12 Capture gamma rays it the copper of the copper-boron layer 0.5 W/cm2
13 Gamma rays from inelastic scattering in both core shelis 4 w/c:m‘2
14 Capture gamme rays in the island core shell 41.4 w/c:m2

 

*In Fig. 30 the data for hecting from scurces 1, 2, 3 are combined and labeled 4,

The heating from sources 10 to 14 was neg-
lected. Their combined contributions to the heat-
ing in the region being considered was estimated
to be about 5% of the total heating. 7

RADIATION HEATING IN YARIOUS REGIONS
OF THE NORTH HEAD

The radiation heating to be expected in various
regions in the north head of the ART was calcu-
lated in order to supply numbers from which ther-
mal-stress calculations could be made. Because
of the complexity and the time that would be in-
volved in calculating accurately the heating in all
the regions of the north head, it was decided to
make preliminary estimates of the deposition rates.
More accurate values calculated for other regions
of the reactor were used as guides. In all cases
the tendency was to overestimate the heating.

Caiculations were made of the heat-deposition
rate in a slab of Incone!l bounded on one side by

 

7 Eor details of these caleulations see Chap, 1.2 of
ANP Quar. Prog. Rep. June 10, 1956, ORNL-2106, p 28.

an infinite fuel region containing the scurces of
radiation. This heat-generation rate was used in
all regions in the north head which are bounded by
finite fuel volumes.

The heat-deposition rates in a slab of Inconel
bounded on one side by slabs of sodium of various
thicknesses were calculated, and the results were
extrapolated and interpolated to obtain the heat-
generation rates in the Inconel regions of the north
head which are bounded by various thicknesses of
sodium.

Fairly accurate calculations were made of the
heat-deposition rates in the Inconel filler plates
below the island ond in the vicinity of the fuei-to-
NaK heat exchanger on the equatorial plane of the
reactor. These results were used as ¢ guide in
estimating the heating in some north-head regions,
and new values were obtained by compensating (by
simple exponential attenuation) for decreased
beryllium thicknesses, penetrations through addi-
tional fuel layers, increased thermal-neutfron leak-
age currents into the north head, etc.

67
 

 

SEGRET
ORNL—LR—DWG 14317

 

 

CALCULATED HEATING {w/cm®)

 

 

 

. _co020 ,-n_,|L,,, -
al _L__ I L o
o

0ca 0.08 Q42 Q4G Q.20 Q24
THICKNESS OF LAYER {(cm]

Fig. 31. Heating in Copper-Beoron Layer by Alpha
Particles from the B'%»,a)Li7 Reacticn.

[t was assumed that a neutron current of 7 x 1073
=1 was escaping uniformly from
the upper portion of the core and that 1 Mw of fis-
sion pcwer was being generated in the fuel regions

neutrons-cm-_ 2-s.ec

of the north head by neutrons escaping into this
region. The latter increased the gamma rays in
the fuel by about 30%.

The sources of gamma rodiation considered were
those from the heat exchangers, the boron, the core
shells, the beryllium, the Inconel sheli capture
gamma rays, the sodium and fuel in the north head,
and the fuel in the core. The sources of heta
particles considered were those from the gases in
the fuel-expansion tank, and the sources of alpha
particles were tcken to be those from boron cap-
The average values of heat generation ob-
in these calculations are presented in

tures.
tained

68

Table 9. The configuration of the north head is
shown in Fig. 32.

BETA- AND GAMMA-RAY ACTIVITY IN THE
FUEL-EXPANSION CHAMBER AND THE
OFF-GAS SYSTEM

The power-source distribution of the activity of
the gases in the space above the fuel in the fuel-
expansion chamber and in the off-gas line has been
determined. The resuits obtained are to be used
in the calculotion of the radiation heating and the
thermal stresses in this region of the reactor,

The radicactive constituents of the gas in this
space will be the gaseous fission products, xenon
and krypton, and their deughter products. There is
also a possibility that some volatile fission-
product fluorides will be formed in the fuel and
will escape into this area. However, it has been
shown '® that if all the fission-product fluorides
entered this space they would add very little ac-
tivity to that already caused by the gaseous fis-
sion products and their daughters. Thus their
effect has been neglected. Also, there is some
question as to whether the daughter products of
the fission gases will actually be carried down-
stream by the off-gas system or whether they will
be deposited on the enclosing walls as they are
formed. In order te get a cecnservative estimate
of the power-source distribution, it was decided to
freat the daughter products of xenon and krypton
as gases (except insofur as their purging from the
fuel into the fuel-expansion chamber is concerned).

The total power and the power density in the gas
space of the fuel-expansion tank as o function of
the volume of the gas and the helium tlow rate are
given in Fig. 33.
the very short- and very long-lived nuclides of
xenon and krypton {along with their decay products)
were neglected, Since the fuel circulation time
in the ART will be less than 3 sec, nuclides with
half lives less than this valve will decay mostly

fn the calculation of the curves

in the fuel before it reaches the purging pumps.
Thus very few atoms with half lives of less than
about 3 sec would get into the gas space. Also,
for nuclides with long half lives {(greater than
chout 100 hr), the number of disintegrations taking

 

18, 1. Newgard, Fission Product Activity and Decay
Heat Distribution in the Circulating Fuel Reactor with

Fission Gas Stripping, TIM-205 (Sept. 28, 1955).

 

 
EECRET
CGRNL-LR—-DWG §4947

  
 

¢ EXPANSION
TANK

—EEN

    

PRESSURE SHELL>/ ,

 

 

 
  
     

 

 

 

 

 

 

 

 

 

 

i
P
LINER—
ey FUEL
0\ T EXPANSION |
TANK
~® /@ i
i
-

 

 

¢

ISLAND

 

 

Fig. 32. Configuration of ART North Head Showing Members Referred to in Table 9.

69
TABLE 9. AVERAGE HEAT GENERATION RATES IN MEMBERS OF ART NORTH HEAD

 

 

Member Heat Generation
No. * Description (w/cms)
1 Pressure shell (below sodium expansion tank} 4
2 Liner 6 w/cm® + 16 w/cm? on expansion-
tank surface due to beta roys
3 Fuel expansion-tank baffle 3
4 Fuel expansion-tank wall 6
5 Upper deck {regions with sodium on both sides) 2
6 Upper deck {regions with fuel on both sides) 15
7 Swirl chamber baffle 3
8 Swirl chomber wall 3
g Lower deck (regions with fuel below and sodium ebove) 8
10 Lower deck (regions with fuel on both sides) 12
1 Copper-boron tiles 25 w/em2t + 6 w/emS, where
t = thickness of tiles (cm)}
12 Fitler block 3
13 Beryilivm support struts 10
14 Filler block 1
15 Copper-baron tile 30
16 Flat section of lower support ring 15
17 Strut part of lower support ring 3
18 Lower support ring 1.5

 

*See Fig. 32 for location of member.

S
GRNL—-_R—DWS 14918

 

3N — e —————— - e e
; POWER DENSITY
| ——— TOTAL POWCR }
20— e . © om0
RELIUM FLOW RATF, 1000 liters /oy (TP}

[ 140

 

—

-

F;. D8 e N I Ce—— = - . b T e o 00

-

i -
& %
- y
- 2
n E
Z 1
o yl
= 2
& O
= =

 

 

o 100 200 302 400 800 600 7OC 820
YOLUME OF CAS 5PACE flf‘..’%)

Fig. 33. Total Power and Power Density in the
Gas Space of the ART as o Function of the Gas
Volume and the Helium Flow Rate for a Fuel
Flow Rate of 22 gpm.

70

place in the fuel-expansion chamber and off-gas
line would be small, since the dwell time at the
assumed helium flow rates is very short. There-
fore these nuciides may be neglected.

In this study 32 nuclides were considered, 16
being isotcpes of xenon and krypton and 16 being
their daughter products. The main contributors to
the power distribution are the daughter products
and not the nuclides of xenon and krypton them-
selves. In all cases the daughter products con-
tribute about 50 to 0% of the total power distribu-
tion. Of the total power, about 90% is due to the
beta-ray decays, with only 10% being due fo
gamma-ray decays. Thus in determining the heat-
ing caused by these gases, it is seen that the heat
deposition will occur mainly in a small surface
layer of the materials surrounding the gases in the
fuel-expansion chamber and the off-gas line.

The power density in the off-gas line as a func-
tion of time ond gas volume for helium flow rates

of 1000 and 3000 liters/day (STP) is given in

 
Fig. 34. The time axis can be converted into
lengths along the off-gas line by dividing the
volume flow rate of the helium gas by the cross-
sectional crea of the off-gas pipe.
gives the power-source density of 1 ecm3 of the

Thus Fig. 34

These plots were made by using the well-known
equations of the decay of parent products and the
buildup of their daughters as a function of time.
The initial conditions at the beginning of the off-
gas line were taken as the equilibrium conditions

gas ot any position in the off-gas line. that weuld prevail in the fuel-expansion tank.

SECRET
ORNL—~LR—DWG t43519

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

    

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

~ T T ‘ I l N T
L[] T T[T T
S~ e e HELIUM FLOW = 3000 liters/day (STP)
5 ~| | ————— HELIUM FLOW = 10CO0 liters/day {STP)
N ' L R
30 g -] S . bl ‘ i
: | GAS VOLUME = 20in.3 | !
i f
| | |
. } ! l
i | |
aEil AN L | L
—_— | | N , | Pl
\ ‘anin3 | ! '
—_ ! e 200 5
"E A \P“T“ 3 i
3 | ‘ 3 T\\ \ |
=20 e e j ] \ ot &
= i l : \ \ _ |
@ | |
= i \ || E
& | N = | |
5 N AU ;
g | T e 3001R.% ! :
e — ] & : L :
——l ‘ ; \\ P :
: - . ! . C
ooin3 | | | T14=.20in ™~ : |
10 t—_-_-_"'i’—""-i' __-__ - fi\_""_\‘ \x i ‘L L
i - \ b.‘-hfi\ \ 1
300in | LT T
- _—"'_"""‘""-!-u—‘»_._w T~ ~~
1 i .--"'-..-.-\ I\E\\ :
i _, 1‘ ™~ ™ _ ‘ . i i |
- | =3 B ; ST T
} M, ’ :\*\ [
; \\ ! \*._:: |
*"l—.____ "
Q i L t e ! '_""‘—v—l
1 2 5 10 2 5 10 2 5 103 2 5 108

TIME {sec)

Fig. 34. Power Density in the Off.Gas Line as a Function of Time and Gas Yolume in the Expansion
Tank for o Fuel Flow Rate of 22 gpm.

71
Part IV

ENGINEERING TEST UNIT
.............

ENGINEERING TEST UNIT

A full-scale zero-power engineering test unit
(ETU) is being fobricated that is a prototype of
the ART. Assembly of the ETU will provide a
test of the feasibility and efficiency of the pro-
cedures prior to assembly of the ART. Thus
time-consuming and expensive rework operations
such as those required for the ARE will be
avoided. The first reactor assembly fabricated
may have, for example, a number of doubtfu! welds
that would not prevent the initiation of nonnuclear
shakedown tests but would have to be reworked
if the assembly were to be used as the high-power
ART. The fabrication experience gained will be
used in assembling the second unit.

A second and even more vital reason for fabri-
cation and cperation of the ETU arises from the
complexity of the stresses on the assembly. The
complex geometry, the wide variety of combi-
nations of thermal and pressure stresses, and the
difficulty of predicting the magnitude and direction
of thermal warping and distortion make it essential
to run a comprehensive test on a nonnuclear
assembly, even though not all the conditions can
be simulated. Experience at the Knolls Atomic
Power Laboratory! has forcefully demonsirated the
importance of such tests when dealing with a
high-performance complex that is to operate under
conditions for which there are little data or ex-
perience. Upcn completion of the tests, the ETU
will be completely disassembled and thoroughly
inspected.

Invaluable shakedown and endurance test ex-
perience will be obtained, and the fraining of the
setup and operating crews during such a test will
expedite assembly and operation of the ART. [t
will be possible, also, to obtain heat transfer
information on the radiators and on the NaK-to-fuel
and sodium-to-NaK circuits and to test some of
the instruments to be used on the ART,

The heat for the nonnuclear ETU will be sup-
plied by two gas furnaces, which will replace the
radiators in two of the four main NaK circuits.
It was originally intended that the capacity of
each furnace would be 5 Mw, but each furnace
has been reduced to a capacity of 1| Mw because
of procurement and instatlation difficulties. Along
with the reduction in the heat input, the radiators

 

TR. W. Lockhart et al., Review of SIR Project Model
Steam Generator Integrity, KAPL-1450 (Nov. 1, 1955).

have been eliminated from the other two man NaK
circuits. Radiators will be included only in the
reflector-moderator {sodium-to-NaK) cooling circuit.
These radiators will permit a determinafion of
performance characteristics of not only the radi-
ators but alsc the louvers and the heat-barrier
doors to be used with them. It is particularly
important that the sensitivity of the controls for
these units be determined at low loads. Since
the units will be essentially the same as those
to be used in the main circuits of the ART, the
test of the reflector-moderator cooling circuit will
serve tc answer meny basic questions regarding
the performance characteristics of the main circuits
on the ART.

Since there will be no after-heat and ro radie-
active off-gases, the fuel dump tank and the
off-gas system, rather than being prototypes of
those to be used with the ART, have been
simplified to expedite fabrication, construction,
installation, and testing of the ETU. The fuel
enrichment, recovery, and sampling systems are
not to be included in this nonnuclear test assembly,
The ART design will be followed in every respect
in the fabrication and assembly of the ETU reactor,
The final reactor for the ART will differ from the
ETU reactor only in modifications brought about
by the fabrication, installation and operation of
the ART. Only those changes considered to be
absolutely essential to the successful operation
of the ART are expected to be made.

SPECIFIC TEST OBJECTIVES

The most valuable information to be obtained
from the ETU will probably be the disclosure of
unanticipated difficulties, such as interferences
in assembly or installation or difficulties arising
from misoperation of certain elements of the
system under peculiar operating conditions, but
several prime objectives have been estabdlished
for the ETU test program. The first and most
important will be a determination of the tendency
of parts to warp, shrink, or otherwise distort during
the original welding and assembly processes and
during testing. Dimensional checks will be made
on all parts during inspection prior to and during
assembly, and the actual dimensions will be
recorded.

The most important sets of diménsions, from the
standpoint of satisfactory reactor operation, are

75
those that offect the moving parts, particularly the
pump impellers, Distortion might cause inter-
ference between the impelier and the stationary
elements of the pump, which, in turn, might cause
malfunctioning of the pump.

Dimensional data on the cooling annuti for the
core, reflector, and pressure shells wiil also be
recorded, |t is important that these annuli be
held to close tolerances, both in local regions
and for the complete assembly. Some types of
deviation will have little effect, while other types
could have quite serious effects. [t is not
possible to state explicitly which ones of the
many possible combinations of devictions from
drawing telerances will be acceptable and which
will not, but it is important to determine the
genercl magnitude and direction of the distortions
in the ETU resulting from assembly operations
and from testing so that the importance of such
distortions in the ART can be reascnably ap-
praised. Unfortunately, the temperature distri-
bution and hence the distorticn pattern in the ART
during high-power operation wifl be different from
those in the ETU, but the distortion during zero-
power operation and during mest of the fow- and
medium-power operation of the ART should be
the same as that in the ETU. Further, many
porticularly bad off-design and fransient conditiens
can be simulated in the ETU, and the effects on
distortion witl be studied,

Flow tests on various components will be carried
out during the assembly of the recctor, These
tests, which moy be made with either water or
air, are partly for checking the calculated pressure
drops through the complex circuits and partly for
calibrating the systems so that they will serve
as ftiewmeters for work during the high-temperature
testing.

WARMUP AND SHAKEDOWN TESTING

During the initiai warmup and shakedown testing
of the ETU a considerable amount of test dota
will be cbtained for use in evaluating the design.
Datc will be taken on the pressure drops through
important elements of the system and, most es-
pecially, on the over-all pressure drop ot each
of a series of given flow rates, and the results
will be checked against the predicted values.
Close attention will be given to the behavier of
the liquid levels in the expansion tanks for the
hot fluids so that the accuracy and the dependa-
bility of the liguid-level indicators caon be de-

76

termined. The heat losses to various elements of
the system will also be measured, and the effects
of operation and position of the heat-barrier doors
and louvers in the auxiliary cooling system will
be determined. It is especially importont that
the sensitivity of the heat losses near the zero-
power condition be determined as a function of
position of both the heat-barrier deors and the
louvers.

OPERATING TESTS

Operation of the ETU should follow the program
prepared for the ART insofar as possible; in
particular, the pump speeds and hence the system
pressures should be programed in the same way
as is planned for the ART. Thus the initial
operation will be carried out af low pump speeds,
the NaK pumps being cperated at probably one-half
speed and the fuel and sodium pumps at abeut
10% speed. After completion of the low- and
intermediote-power simulation in the ETU, the
pump speeds wiil be increased to full design
operating values. During this simulated operation
it will be possible fo obtain further test data on
the fiow characteristics of the various systems,
the heat balance data, and some indication as
to the performance of the heat exchangers, par-
ticularly those in the reflector-moderator cooling
circuit. The precision with which the heat balance
data can be obtained will be determined by
checking the heat balance detc for the air, the
NaK, and the sodium systems against each other.
During the shakedown operations it will probably
be desirable to determine the effect that cutting
out one or more pumps will have on the liguid
levels in the expansicn tanks. Also, it may be
possible to conduct tests on the performance of
the xenon-removal system,

A carefuily programed series of thermal-strain-
cycling tests will be included in the ETU oper-
ational tests, Delineation of this program will
be delayed until the stress analysis work is
essentially completed so that significant tests
can be made. Hence the precise temperature
levels, pressure levels, and flow rates that should
be used in this program connct be specified until

around December 1954.

REACTOR ASSEMBLY

The reactor is made up of five major subase
semblies: the reflector-moderater, the main heat
exchanger, the north head, the isiand and south

 
- e e e L e

 

 

pressure-shell liner assembly, and the pressure
shell. Each of these major secticns is to be
assembled and then fitted together in the proper
sequence fo produce a complete assembly,

Many of the individual components of the subas-
semblies present difficult fabricational problems
because of their geometric shape and the di-
mensional tolerances specified. A particularly
difficult problem will be encountered in helding
close tclerances on complicated weldments and
on the thin-walled concentric shells which form
the fluid passages and separate the components
of the reactor. The most meticulous and rigorous
inspection techniques available are to be used
on materiatls and welds.

Assembly of the reacter will start with the
reflector-moderator, which is made up of the outer
nconel core shell, the beryllium hemispheres,
the strutering structure, the B,C layer, and the
inconel reflecter shells., The upper and lower
halves of the Inconel outer core shell will be
welded together at the equator, and the upper
coliar will be welded to the top of the shell. The
Inconel spacers will be fitted on the inside surface
of the beryllium hemispheres, and the beryllium
will be fitted around the Inconel shell, The
spacers on the outer surface of the beryllivm will
then be installed, together with the canned copper-
B,C patches, at the north end. The strut-ring
assembly and the Inconel reflector shell that
houses the beryliium wili then be welded together
and to the outer core shell. Next, the B,C tiles
wiil be positioned on the outer surface of the
assembly and covered with the 1/1{ -in.=thick Inconel
shell that serves as the boron jacket. The B4C
tiles are to be placed in sheet metal containers,
one half of which will be spot-welded to the
surface of the shells and the other half will be
slipped into the attached helf fo form a container
around the tile. This operation will complete the
reflector-moderator assembly.

Special movable fixtures will be used to place
the 12 tube bundles of the main heat exchanger
arcund the reflector assembly., The units must
be fitted into place simultanecusly and held in
posifion s¢ that the north-head assembly may be
lowered over the reflector-moderator—heat ex-
changer assembly. The north head is o compli-
cated weldment containing the fuel pump volutes
and housings, the fuel-expansion tank, the core

entrance header, the sodium pump volutes, and
the seodium-to-NaK heat exchangers. The north
head will be built up from subweldments of the
pump velutes and header passages on two decks.
Welding accessibility and welding seguerice to
prevent excessive warpage of critical surfaces are
the most difficult problems envisioned for this
assembly at the present fime. Weldability medels
that illustrate the steps involved in assembling
the north head are shown in Figs. 35 through 41.

The island and south pressure-shell liner as-
sembly will be assembled by fitting the upper and
lower beryllium sections together and placing the
spacers on the beryllium surface. The upper and
lower sections of the inner core shell will then
be placed around the beryllium, and the equatoricl
weld will be made. The upper island and the
expansicn joint will be welded to the core shell
to form the istand assembly, The south pressure-
shell liner assembly will then be assembled with
the shells containing the neutron shieldirg and
welded to the island assembly. The islond will
be inserted through the moderator assembly so
that the bellows assembly will slip into positicn
in the north head and sc that the southern pressure-
shell liner will seat against the northern section
at the equater. The equaterial weld will then be
made,

The upper half of the pressure shell will be
lowered over the reactor assembly, and the lower
pressure shell, with the laminated filter plates
in place, will be brought into position. The girth
weld will be made for joining the two halves of
the pressure shell. The upper island connection
will be welded to the upper pressure shell, and
the heat exchanger header pipes will be welded to
the pressure shell sleeves.

The sodium expansion tank will be welded to
the upper pressure shell, and the control rod
sleeve will be welded at the top of the expansion
tank to complete the reactor assembly.

The assembly will include the lead shield in
order to obtain a test of the support structure ond
the cooling systems, The water shield will be
omitted on the ETU fo ease procurement and
installation problems. The many shield pene-
trations for instrumentation, helium, off-gas, and
other connections to the reactor and reactor shell
appurfenances make the detailed shieid design and
instatlation very difficult,

77
 

Fig. 35. North-Head Weldability Mode! Showing Lower Deck and Peripheral Ring, Step 1.

Na -~ TO - NeK o
HEAT EXCHANGER BUNDLE

 

Fig. 36. North-Head Weldebility Medel, Step 2.

78

 

 
g

 

b apeweT
PHOTO 25543

“’!l?-x_.i_ .
f‘“ TOP DECK OVER Ng~TO~NaK

EAT EXCHANGER

Fig, 38, North-Head Weldability Model, Step 4.

 

79
80

|

FUEL PUMP VOLUTES

 

F FUEL EXPANSION
TANK WALL

TOP DECK OVER
FUEL VOLUTE

Fig. 40. North-Head Weldability Model, Step 6.

P
PHOTO 25539

Ihc puve)

[ PO
| PrioTO 25542

 

 

 
 

 

PROTO 25540

Fig. 41. North-Head Weldability Model, Showing Another View of Step 6.

REACTOR DISASSEMBLY

Dimensional data will be taken at frequent
intervals throughout the disassembly of the ETU.
The first step after removal of the reactor from
the test stand will be the determination of the key
dimensions of the pressure shell. The over-all
heights from the level of the pump mounting
flanges to the north head and to the south head
will be measured. A cut will then be made at the
equator through both the pressure shell and the
pressure-shell liner. A cut will also be taken
through the thermal sleeves around the heat ex
changer header outlet tubes at the south end, and
the sleeve that attaches the island assembly to
the north head will be severed. This will permit
the removal of the island and the southern half
of the pressure shell. The core shells, pressure
shell, and the heat exchanger can then be ex-
amined. Dimensional data on the pump wells,
on the pump impellers, and on the areas in the
vicinity of close impeller clearances will be

taken. The welds that attach the pressure shell
dome to the north head around the roots of the
fuel pump barrels will be milled out, arnd the
transfer tubes that attach the reflector assembly
to the north head, together with the thermal
sleeves and heat exchanger outlet tubes at the
north head, will be cut. The twe regions can then
be separated for inspection. The reflector shell
will be cut, and the beryllium will be removed and
inspected for corrosion, spacer fretting, and
thermal cracking,

Other elements of the ETU system, including
the snow traps, the celd traps, the filters in the
NaK system, efc., wiil also be inspected. Typical
sections of piping will be removed and inspected
for corrosion and mass fransfer. The NaK pumps
will be checked for changes in impeller running
clearance, the dump valves will be inspected,
and all elements of the plumbing will be carefully
dye-checked for thermal cracks that might have
been induced during the cperation.

81
Part V

CONSTRUCTION AND OPERATION
 

 

CONSTRUCTION AND OPERATIOR

PLANS FOR INSTALLATION OF THE ART

The present plans for assembly and installation
of the ART were evolved to minimize the time re-
quired to get the ART intc operation. It was
recognized that not only must the ART assembly
start before completion of the ETU tests but also
that operation of the ART would have to be de-
bayed if operation or disassembly of the ETU
indicated the need for modifications of the reactor
unit. As a result, three complete reactor units are
being ordered. Subcssemblies of the third unit
will be storted as soon os the second unit is
assembied, but fina! assembly will not be started.
The second reactor will be assembled and installed
in Building 7503 as rapidly as possible and, if no
trouble develops in the ETU, will be ploced in
operation as soon as the disassembly and the
inspection of the ETU have been completed. How-
ever, if disassembly of the ETU indicates the
necessity for modifications, work will be started
en the third set of subassemblies as scon as
decisiens are reached as to the nature of the modi-
fications required. Final assembly of the third
unit would proceed concurrently with the removal
of the secend unit.

OPERATION OF THE ART

Filling eand Heating

The initial fiiling and heating of the reacter are
part of both the testing procedure and the operating
procedure by virtue of the reactor being of the
circulating-fuel type. The NaK system will be
filled with NaK, ond, then, with the heat-barrier
doors closed, the NoK pumps will be started. The
power provided by pumping the NaK at one-half
design-point flow will supply most of the heat re-
quired, and electrical heaters will supply the rest
of the heat needed to bring the system up to 1200°F
in @ minimum of 24 hr,

The sodium for cooling the reflector-moderator
and the sodium for cocling the control rod will be
added during the heating peried when the system
temperature is about 350°F. With the sodium pumps
operating, it will then be permissible to add more
heat to the system through the main NaK system.
With the system isothermal at 1200°F, the cold-
trap systems will be gradually cooled down in
order fo remove oxides from the NaK,

The fused-salt fuel carrier will be put into the
fill-and-drain tank when the entire system is in
an isothermal condition ot 1200°F. The fuel pump
will be started and will be operated at a nominal
speed of 100 rpm, ond the fuel carrier will be
pressurized info the reactor and heat exchanger
system, The dump valves will then be closed,
and the fuel and sodium pump speeds will be ine
creased in order to degas the systems, When the
system has been degassed and the design fuel
has been obtained in the swirl tank, the
main NaK pump speeds will be increased to design
point.
will be made stepwise in about six steps from one-
half speed to design-point speed.

level

Increases in the main NaK pump speeds

Isothermal operation with all pumps operaring at
design-point speed will be maintained for about
24 hr. With all the pumps operating at design point
the added power will raise the temperature of the
system. The electrical heating will be decreased,
as required, in order fo maintain the system in an
isothermal condition. If it is necessary, a main
blower will be started and the heat removal will be
controlled by manipulation of the auxiliary louver
positions so that the system can be maintcined in
an isothermal condition with all pump speeds at
the values desired,

After no more than 24 hr of isothermal operation
of the system, the dump valves will be opened and
the fuel carrier will be dumped into the fili-and-
drain tank. Opening the dump valves will autc-
matically reduce the main NaK pump speeds to one-
half design=point speed and at the same time will
reduce the fuel and sedium pump speeds tc estab-
lished minimums, '

Samples of the fuel carrier, the NaK, and the
sodium will then be taken and examined for corro-
sion products. Mass spectrographic analyses of
these samples will also be made in order to de-
termine whether any leakage occurred between any
two adjacent systems,

[f there is no evidence of leakage and if the
corrosion-product analysis indicates that fthe
carrier is clean, a portion of the carrier will be
withdrawn from the fiil-and-drain tank to moke room
for the addition, in two steps, of sufficient N02UF
to provide a fuel mixture which hos about 40% of
the U235 required to achieve criticality with the
reguloting rod withdrawn., The estimate of 60% is

85
based on the system volume and on the data ac-
quired during the high-temperature critical experi-
ments described in Part H of this report.

Enriching to Critical

After addition of the initial charge of Na UF,,
mixing will be accomplished by pressurizing the
fluid into the system until its surface level is
approximately at the mid-point of the reactor; the
pressure will then be released 1o cllow the fluid
to return to the fill-and<drcin tank. It is estimated
that five or six such cycles will be required to
produce adequate mixing. During this mixing pre.
cedure the control rod will be at a withdrawn posi-
tion of about one-fifth its full stroke.

The fuel mixture will be pressurized to fill the
reactor system to the operating fluid level. This
operation must proceed slowly as it nears com-
pietion, [n order to avoid overfilling the reactor, a
level indicator will be provided in the fuel overflow
line, in addition to the level indicator in the fuel-
expansion tank. The pressure in the dump tank
witl be limited by a pressure-relief regulator to a
value that will be just sufficient to force the fuel
up to the minimum level. This pressure will have
been estimated from the fluid shakedown tests.

I proper mixing has been accomplished, addition
of the initial charge is not expected to bring the
reactor to criticality, since the fuel addition was
calculated tc give o reactivity value () of about
0.94. Improper mixing could, however, produce
seme abnormal effects, since the fuel pumps will
be operating at the established minimum speed.
The slowly circulating fuel could pass through
the core in an alternately rich and lean stream,
and the count rate of the fission chambers would
follow these patterns of fuel density. Since design-
point circulafion cccurs only when the fluid fevel
is high enough for the pumps to do some pumping,
pauses in the mixing cycie cannot be expected to
improve the mixing.

During the pressurizing of the initial charge of
fuel into the reactor, safety instrumentation will
be set on either a fast period or an estcblished
maximum neutron flux level to dump the fuel by
avtomatically releasing the pressure on the fill-
and-drain tank, since the fuel drain valves will
not have been closed. The operation will not be
carried out with full reliance on these safety
devices. As is customary for a new reactor, much
of the operation at this stage will depend upon the
skill of the frained and experienced operator,

Fast regulating=rod insertion af the calculated
rate of change in Ak/k of -—-E/é% per second and
concurrent release of the helium pressure over the
fill-and-drain tank will provide the means for re-
versing a fast rate of rise of the neutron flux.
These operations, in addition to being autematically
actuated, con be manually set in motien by one
manual switch on the consecle in the control room.

It is believed that poor mixing would cause the
filling of the reactor, after the addition of the initial
charge of enrichment mixture, to be the most haz-
ardous step in the enrichment procedure. Ex-
perierice with the high-temperature critica! assembly
indicated, however, that adequate mixing can be
achieved. Successive charges of enrichment
material will be so small thet no one charge could
bring the reactor to criticality unintentionally, even
with the poorest of mixing. The charges of entiched
material made after the initial one will be in quan-
tities that will increase the reactivity by an average
amount of about 0.3% for a given rod setting.

After the initial enrichment charge has been
oroperly mixed with the carrier and the mixture has
been pressurized into the fuel system of the re-
actor, count rates will be taken for @ minimum of
five positions of the reguliating rod., These posi-
tions con be determined accurately aond will be
recorded,

The fuel will then be dumped into the fiil-and-
drain tank, and another accurately measured en-
richment charge will be added. The mixing opera-
tions will be repeated. The fuel will again be
pressurized into the system, and count rates will
again be tcken for the same positions of the regu-
lating rod and will be recorded. This procedure
will be repeated for all successive additions of
enriched material. Curves showing the reciprocal
of the count rate as a function of the mass of U235
in the system will be plotted for each of the five
rod positions after each addition is completed.

At some stage in the enrichment procedure, be-
fore the reactor is critical with the rod withdrawn,
the fuel dump valves will be ciosed to permit an
increase in fuel and sodium pump speeds. These
pump speeds will be brought tc design point, and
the heat removed in the auxiliary system will be
adjusted to maintain the system in an isothermal
condition. The count rate will be taken for each
of the five rod positicns cfter the pump speeds
have been increased.

With fuel and sodium pumps operating at their
respective design-point speeds and with the main

 
 

..........

NakK pumps operating ot one-half design-point
speed, a temperature differential will be imposed
on the system by opening the heat-barrier doors,
starting a blower, if necessary, and opening the
main fouvers a small amount. The exact conditions
to be used will have been determined in the earlier
fluid shakedown tests. This operation will be
carried out with the reguicting rod inserted fo
about one-fifth its full fravel.

The mean fuel temperature will be continuously
recorded, and the count rate will be recorded. The
relation between the count rate and the mean fuel
temperature should give semiquantitative evalua-
tion of the fuel temperature coefficient of reactivity.
It will be necessary to ceol the NaK system slowly
and tc have completed enough of the enriching to
get o sufficiently high count rate te obtain reliable
statistics.

After these date have been cbtained, the heat
removal will be stopped; that is, the biower wiil be
shut off, the louvers will be closed, oand the heat-
barrier doors will be closed. The system wili be
returned to the previous isothermal condition at e
temperature of 1200°F by the operating heaters and
the pumps as described previously. The fuel and
sodium pump speeds will then be reduced to the
established minimums, and the fuel will be dumped
into the fill-and-drain tonk.

Successive additions of Na,UF, wiil be made
until the reactor is critical with the regulating rod
fully withdrawn and then until the reactor is critical
with the rod 80% withdrawn. Inhour curves will
be obtained from data taken at each rod position at
which the system is critical, in the usual way.
The data will be obtained with the fuel pumps
operating at the established minimum speed. Under
these conditions the reactor may be considered as
a stationary-fuel reactor for the purpose of fixing
the fission-fragment delayed-neutron yield.

The delayed-neutron contribution to reactivity as
a function of fuel pump speeds can best be deter-
mined in the low-power-level experiments by
holding the reactor critical and at a constant flux
for all pump speeds. This will be done by the flux
servo system, which will hold the flux constant
by withdrawing or inserting the regulating rod upon
an increase or a decrease of the pump speeds. By
the time that this experiment is conducted the regu-
iating rod will have been calibrated for the sta-
tionary-fuel case, or, if not, the data will be in-
terpreted after the rod has been calibrated.

Low-Powerstl.evel Experiments

Low-power-level experiments will be completed
before all the enriching has been done, that is,
when sufficient enriching has been done to bring
the reactor critical with the regulating rod 80%
withdrawn,

The reactor will be brought critical when the
system is isothermal at 1200°F, The dump valves
will be closed, and the neutron flux will be set at
a nomina! value estimated tc represent a nuclear
power of 10 w, Control of the rod will be placed
on the flux servo, with the fuel and sodium pumps
operating at their established minimum speeds,
Both the fuel and the sodium pump speeds will
then be gradually brought up to design point, The
position of the regulating rod witl be continuously
recorded, as will the pump speeds. The servo
system will withdraw the reguiating rod t¢ main-
tain constant flux,

The speeds of both the fuel and the sodium
pumps witl be decreased slowly to their established
minimums while the servo holds the flux constant.
This operation will be carried cut carefully and
probably at a slower rate of decrease in pump
speed then was possible for the rate of increase,
since, at some point, the rate of decrease in pump
speed will be such that the servo cannot keep the
flux from rising. Abruptly stopping ¢ fuel pump
will automatically initiate a fast regulating-rod
insertion at a rate of change in Ak/k of—%% per
second. These experiments should provide the
data reguired for evaluating the delayed-aeutron
contribution to reactivity as o function of the fuel
pump speed.

With the regulating rod controlled by the servo,
the flux will be held constant and the sodium and
fuel pumps will again be brought up to design-point
speeds. The system will be made isothermal at
1200°F and then cooled, at a rate previously de-
termined in the fluid shakedown tests, by opening
the main barrier doors and the main louvers, as
necessary.

The mean fuel and sedium temperatures and the
rod position will be continuously recorded. The
cooling of the system should cause the rod to be
inserted and permit a quontitative evaluation of
the negative temperature coefficient, Part of the
temperature coefficient will be derived by zooling
the fuel and part by cooling the moderator. Careful
analysis of the rod position, and, accordingly, the
Ak/k, vs the fuel mean temperature will, however,
give a reliable estimate of the fuel temperature

87
coefficient of reactivity. This estimate will be
used in conjunction with the determination of the
over-all sysftem temperature coefficient and the
sodium temperature to estimate the moderator tem-
perature coefficient of reactivity. If the rod cali-
bration at this stage is thought to be inadequate,
the data from this experiment can be saved until
the rod calibration has been completed, and then
these coefficients can be estimated in terms of
(Ak/R)/°F.

After completion of these experiments the system
will be brought isothermal at 1200°F at the con-
stant filux with the regulating rod on servo controi,
Then the regulating rod will be inserted manually,
the fuel and sedium pump speeds reduced to their
established minimums, and the fuel dumped into
the filt-and-drain tank.

The enrichment precess will then be resumed,
with the same procedure being used as that de-
scribed above, until the reactor is critical with
approximately 20% of the regulating rod withdrawn.
Spot checks of the delayed-neutron contribution to
the reactivity as a function of the fuel pump speeds
may be made before complete enrichment has been
effected. Likewise, spot checks of the tempera-
ture coefficient of reactivity moy be made before
enrichment has been completed.

Operation at Power

A final low-power-level check of the temperature
coefficient of reactivity will be made before opera-
tion at power is undertaken. This check will be
made by the procedure described above.

The reactor will then be taken to power in the
same sequence as thot used for the ARE. With
the system iscthermal at 1200°F and critical, the
regulating rod will be withdrawn until the reactor
is on o positive period of about 20 sec. For o
while the log N recorder should trace a straight
line. The line will be straight as long as the
positive period is constant. As soon as the nuclear
power is high encugh to heat the fuel appreciably,
the periocd will increase because of the negative
fuel temperature coetficient of reactivity. Like-
wise, the slope of the log N curve wiil decrease.
At the time that the period starts increasing, the
main and auxiliary barrier doors will be opened.
Opening these barrier doors will put _some load on
the reactor. The log N curve should then level off,
either at a higher or lower level than that existing
when the barrier doors started to open or at the

88

same level, Since the barrier doors can be main-
tained entirely opened or entirely closed, there
will be no means for perturbing the load at this
stage of the operation. In fact, it will be advisable
not to attempt to keep the power at this level too
tong, since there will be no means for adequately
controlling the varicus temperatures. The mederator
may overcool or overheat,

With the louvers closed, @ main blower will be
started. This will cause a flow of air across the
radiators and will increase the power extracted
from the reactor. Without any regulating rod adjust-
ment, the log N should increase and level off at a
new level. At the new level, as a result of the
negative fuel temperature coefficient, the system
stability should be comparable to that of the ARE
with the heat-barrier doors open. For the ARE the
foad was about 200 kw., For the ART, comparable
stability should exist at cbout 1.2 Mw. The
proeduct of the fuel heat capacity in the reactor
and the fuel temperature coefficient for the ART
should be approximately six times the corresponding
product for the ARE, These factors determine the
power level at which cemparable stabilities may
be expected. At 1.2 Mw the reactor power shouid
level out promptly without excessive ‘‘ringing'’
on small perturbations in load or rod position.

it will be advisable to open the main louvers
gradually until the nuclear power is from 6 to 10 Mw,
Perturbations in foad and in rod position can be
made at this power range, and the sodium tempera-
ture control can be maintained. A power run ex-
tending for several days can be made then with a
mean fuel temperature of 1200°F and o maximum
power output of just over 23 Mw. After this run
the reacter will be shut down by slowly closing
the main NaK louvers and, finelly, by inserting the
regulating rod.

The fuel will then be dumped into the fill-and-
drain tank, ond a sample will be pressurized into
one of the sample cellis and sealed off with 4
bismuth freeze valve. The fuel will again be pres-
surized info the system, and the reactor will be
taken critical while isothermal at 1200°F. To
maintain the system isothermal will require a
greater air load than was needed for the clean
system because of the afterheat frem fission
fragments.

With the reguicting rod on servo the fuel and
sedium pumps will be slowly brought to design
peint, and xenon purging will be allowed to proceed
until the rod position has become constant for
several minutes, Then the helium bleed into the
purging system will be decreased for several
minutes fc determine its influence on the xenon
purging. A decrease in purging should cause the
regulating rod to withdraw slowly. This stage in
the experiment will be the most faverable for xenon-
purging tests, since the system will be free from
temperature coefficient compensations on perturba-
tions in xenon removal and since the iodine
precursor of xenon can be evaluated accurately as
a function of time. The only xenon removal which
can occur at large rates is that provided by the
purging system.

After experiments to determine the character-
istics of the xenon-purging system have been com-
pieted, the reactor will be taken to its maximum
power with a 1200°F mean fuel temperature {approx-
imately 23 Mw). The regulating rod will then be
withdrawn until the mean fuel temperature in the
reactor is 1250°F. The louvers will be opened
until the NaK temperatures at the radiator outlets
are 1070°F. This will give the maximum power
loed available with @ mean fuel temperature of
1250°F., This power level will be maintained for
several days and then lowered by inserting the
reguiating rod until the mean fuel temperature is
1200°F, Then the shutters will be closed, the
fuel will be dumped, and another sample will be
taken.

Procedures for operating the reactor at mean fuel
temperatures of 1300, 1350, and 1400°F will be
precisely the same as those described for operation
at 1250°F, Each power run will continve for
several days, with the condition imposed that the
duration of all power runs at or above 20 Mw shall
not exceed 1000 hr, After each power run the fuel
wili be dumped and a sample will be taken. Af
the end of the last power run the fuel will be
dumped and will be cooled in the fill-and-drain
tank until the ofterheat has decreased to such an
extent that the fuel can be pressurized into the
recovery tank.

PLANS FOR DISASSEMBLY

it has been recognized from the inception of the
ART project that the disassembly of the reactor
would be o very difficult operation. It is, of
course, extremely important that a fairly complete
disassembly and ‘‘post-mortem’’ be conducted on
the reactor to determine how well it withstood the

test conditions., Information will be desired on

corrosion, deposits, plating-out of fission products,
structural stability, distortion, warping, cracking
and local behavior in the vicinity of welds, brazed
joints, etc., compatibility of beryllium and Inconel,
integrity of the boron-carbide layer and its cladding,
and radiation damage effects on the various ma-
terials, particelarly those associated with the
boron-carbide layer, etc.

Enough information is now available to present a
fairly comprehensive picture of the disassembly
problem. It seems best that a rough disassembly
be carried out in a hot cell designed for the purpose,
The heavy disassembly operations can be carried
out in this cell, and the small samples can be
removed in lead pigs and examined in smaller
specialized hot cells well removed from the very
large amounts of activity associated with the main
reactor assembly.

The first step in the analysis of the disassembly
problem has been the determination of source
strengths for the various activities. Basically,
there will be, of course, the activity of droplets
of fuel that will adhere to the surfaces of the fuel
circuite Similarly, there wiil be some activity from
sodium droplets that will odhere to surfaces of
the sodium circuit.  The Inconel will kecome
activated, primarily the cobalt constituent, as will
the beryllium, both because of the presence of
cobalt impurities and scandium impurities. In
addition, radicactive elements will be plated-out
from the fuel, namely, ruthenium, columbium, and
molybdenum.,  Further, it can be expected that
fission-fragment activity will be buried in the
Inconel core shells. While both the adhering and
the plated activity might be removed by a drastic
cleaning operation, such as an acid etch, such a
cleaning operaticn is obviously undesirable be-
cause it would destroy the valuable evidence
wanted for corrosion analyses. While it is difficult
to estimate the amount of liquid that wili adhere
to wetted surfaces, it cppears that roughly 2% of
the liquid in both the fuel and the sodium circuits
will remain in the reactor after it is drained. A
fiushing operation carried out with a non-uraniume
bearing fluoride mixture will remove probably 90%
of this residual activity, However, such a flushing
operation will not remove plated-cut activity or the
activity of the fission fragments buried in the core
shells, The principal advantages of the flushing
operation would be o reduction in air-borne con-
tamination during disassembiy, o probable reduction
in the activity of parts wetted by fuel in the lower

89
temperature portions of the system, a marked re-
duction in activify held up in fhe thermal sleeves
at the south end, and a very marked reduction in
activity in the drain valves and the drain line.
The problems that would be posed by the activity
in the drain fine ond valves during the removal of
the reactor from the cell might alone justify o
fiushing operation. Unforfunately, this operation
would have tc be carried out after the fuel had
been transferred from the fill-and-drain tank to the
fuel recovery tank.

Estimates of the activities associated with the
various parts have been made for 100 and 300 days
after shutdown and are tabulated in Table 10. The
majority of the data are from decay curves prepared
by Bertini.! From the table it may be seen that
the principal sources of activity are the fission
oroducts in the fuel, the plated-cut activity (princi-
pally ruthenium), the cobalt in the Inconel core

 

TH, W, Bertini, ANP Quar. Prog, Rep. March [0, 1955,
ORNL- 1864, p 24.

shells, and the scandium and cobalt in the beryl-
tium. Each one appears to give a source of the
order of 1,000 te 100,000 curies 100 days after
shutdown, All the other activities seem to be of
the order of 10 curies or less; therefore if plating-
out were not a problem, it should be possible to
remove smcll specimens from the reactor and for
each specimen to have an activity of less than
1 curie. It is evident that, except for the fuel
(which decays by a factor of 10 in the period from
100 to 300 days after shutdown) and the plated-out
ruthenium, the activities are long lived and that
very litile advantoge would be geined from waiting
the 300 days, except for parts on which plated-ocut
activity should be expected. While little informa-
tion is available, it appears that the bulk of the
plated-out activity will be fairly well distributed
the core shells, the heat exchanger, and
other surfaces of the system wetted by the fuel.

The only isotopes of the plated-cut elements that

need to be considered at times long after shutdown
(greater than 100 days) are Nb93, Ru'93, and

over

TABLE 10. ESTIMATED ACTIVITY OF ART COMPONENTS AFTER 500 he AT 60 Mw

 

Activity {curies)

 

 

Source 100 Days After 300 Days After
Shutdown Shutdown

Cemplete chorge of fuel 10¢ 105

% of fuel charge 164 1000
Plated-out activity {total activity of Ru, Mo, and Nb in fission products) 2% 107 8000
P lated-out activity near core outlet, per in.? 1 0.05
Complete charge of No 10 7
2% of Na charge 0.2 0.14
Complete charge of NaK 0.2 0.2
2% of NaK charge 0.004 0.004
Inconel core shells {not including adhering or plated-out activity} 7000 5000
Pressure shell {not including adhering or plated-out activity) 40 30
Heat exchanger tubes {not including adhering or plated-out activity) 40 30
Berylilium 900 700
Leed shield G.1
Specific activity of lnconel eutside reflector boron layer near equator, per g 10—4 104
Specific activity of Inconel outside reflector boron layer in north or south 10=3 10~3

heads, per g
Droin valves and line from reactor to dump tank, 2% of fuel volume 300 30

 

90
Rul®, All the other isotopes have sufficiently
short half lives that they will have decayed
appreciably in 100 days and thus may be neglected.
The only daughter products that will have large
activities af shutdown times greater than 100 days
will be Rh'%3™ gnd Rh'06 (daughters of Ru'9?
and Rumé‘ respectively).

The activities, in curies, at shutdown and ot 100
and 300 days after shutdown for reactor operating
times of 500 and 1000 hr ot 60 Mw are given in
Table 11. Also given are the cverage beta-ray
energy and the average gamma-ray energy per dis-
integration. [t has been assumed that, at shutdown,
the fuel will be dumped, and so ne more plating-out
in the reactor will take place. While the totcl
plated-out activity is high when expressed in
curies, the bulk of it, as can be seen in Table 11,
is in relatively soft- gamma rays and the con-
tribution to the radiction dose outside the pressure
shell is really essentially that from the Nb®5 (or
less than 10% of the total activity in curies) and
even that gives only a 0.75-Mev gamma ray, which
is aftenuated 2.5 times as rapidly as the hard
fission-product decay gamma rays.

In extrapolating experience that was gained in
the disassembly of the ARE, the differences in
proposed power level and operating time for the
ART were taken into account. The ARE was
operated for o total of 100 Mwhr, whereas the ART
is scheduled to operate for a total of 30,000 Mwhr.
This difference will mean vastly increased dif-
ficulty in disassembly. Since the bulk of the
activity 100 days or more after shutdown will be

from long-lived isotopes, the activity will be
directly proportional to the number of megawatt-
houts that the reactor is operated.

The information in Table 12 was prepared on the
basis of that in Table 10 to show the effective
source strength of the reactor at various stages of
disassembly. |t is evident from Table 12 that work
can be carried out immediately adjacent to the
exterior surface of the water shield 100 days after
shutdown without the need for auxiliary shielding.
Some work can still be carried out immediately
adjacent to the reactor after removal of the water
shield, but this work will have to be limited to a
matter of 10 min or so. After removal of the lead
shield, all work wiil have to be done remotely, A
possible exception might be removal of the lead
shield from the hot cell. The thickness of the
lead shield required for this operation weuld be of
the order of 3 in. It is not yet clear whether the
space avoilable in the hot cell will make removal
of this lead shield essential. _

The limits on the source activity for various
types of hot laboratory work are indicated in
Table 13. Stondard laboratory tolerance dose
rates for operating persennel were presumed in
preparation of this table. As a first step in es-
tablishing safe werking conditions, an attempt will
be made to get a good separation distance between
the operator and the source so that it will be
possible to work with sources which are on the
order of 0.1 curie. The next step will be the use
of a shadow shield barrier to cut out line-of-sight
radiation and tc employ mirrors for viewing hendling

TABLE 11. ACTIVITY OF PLATED-QUT MATERIALS IN THE ART AFTER 500 AND 1000 hr OF OPERATION AT 60 Mw

 

Activity at Shutdown Activity 100 Doys After

 

 

 

Average  Average . Activity 300 Days After
Nuclide Bete-Ray  Gamma-Ray (?Zmz;‘zog;"]fi; (curies) Shutdown (curies) Shutdown (curies)
Energy Energy i egrations)  After SO0 hr - After 1000 hr  Atier 500 hr  After 1000 hr  After 500 hr  Aiter 1030 hr
{Mev) (Mev) of Operation of Operation of Operation of Operation of Operation of Operation
NBIS 0,053 0.745 100 1,29 x 105 4,08 x 105  1.82x 104  565x 104  336x102  1.06x10%
Ry103 0.074 0.498 99 565x 105 9,57 105  LO0Ox 10 1.68x10°  3.06x10% 51sx103
Ry106 0.0131 0 0 7.84 x 108 153 104 6.49x 108 K26 x 104 4.44x10°  8.65x 103
Ryl03= ¢ 0.04 100 0 0 9,20 x 104 L60x 10°  2.92x10°  4.92x 103
Rh106 1.05 0.513 25 0 0 6.49 x 103 1,26 104 4.44x10°  8.45x 10®
0.624 12
0,87 1
1.045 2
1.55 0.5
2.41 0.25

 

 

21
 

TABLE 12, EFFECTIVE SOURCE STRENGTH OF REACTOR AT VARIOUS STAGES OF DISASSEMBLY

 

Equivalent* Source Strength

 

 

Weight of {curies)
Ce tor A bi C t
mpenent or Assembly emponen 100 Doys After 300 Days After
or Assembly
Shutdown Shutdown .
Reactor with complete lecd ond water shield 85,000 1 ¢.1
{without Na or fuei)
Reactor with lead shield only 43,000 10 2
Reactor and lead shield with pump bedies removed 41,400 20 40
Fuel pump body assembly 400 10 7
Na pump body assembly 400 1 1
Reactor without lead shield or pump bodies 11,200 1000 100
Total estimated activity in 50 b of chips from cuts 50 10 7
{except core shells and heat exchanger cutlet)
ZI'F4 vapor trap (1% of activity in fuel) 150 5000 500
Control rod 20 70 50
Total estimated activity in 5 lb of chips from cuts 5 100 3

through core shells and other regions containing

high concentration of plated-out activity

 

*The ‘*equivalent source strength’’ is taken to be a 2-Mev gamma-ray source emitting photons at the same rate as
the assembly in question. The radiction dose thus would be roughly 1 t/hr curie at 1 meter from the center of such

g source and would vary inversely with the distance from the center of this equivalent source.

TABLE 13. LIMITS ON SOURCE ACTIVITY FOR HOT LABORATORY WORK

 

 

Operater-Source
. . Maximum Activity of Source
Condition Separation Distance*
{curies of Z-Mev gamma rays)
(1)
No shielding i5 0.2
No shielding 30 0.8
Shadow shield for line-of-sight radiation 15 10
{20 in. of special concrete, density = 3.0 g/cms,
or 4 in. of Pb)
Shadow shield for line-of-sight radiation 30 40
(20 in. of special concrete, density = 3.0 g/cms,
or 4 in. of Pb}
Enclosed hot cell (40 in. of special concrete or 15 500,000

Zn sz, density = 3.0 g/cm3, or 10 in. of Ph)

 

*A permissible dose rate for the operator of 8 me/hr is presumed.

 

92
 

operations. |f these steps are taken, a source
sirength approximately 30 times greater than would
be acceptable without shielding con be tolerated.
Finally, it is possible to use a completely en-
closed, heavily shielded hot cell. If this is to be
done, it appears that work on the ART would require
a hotecell wall thickness of 48 in. of special con-
crete of zinc bromide having a density of 3 g/cm?,
or the equivalent of 10 in. of lead,
It is important that all the infoermation provided in
Tobles 12 and 13 be considered to be tentative,
The activities are based on the premises indicated,
the most important being that 98% of the fuel will
drain. This analysis was prepared in order to give
a fair over-all picture of the activities with which
it will be necessary to work.
[Listed below are the major steps that will be
involved in removing the reactor from the reactor
cell and moving it to the hot cell, as well as dis-
assembly operations that can be carried out in the
hot cell prior to closing the top if it is essential
to remave the lead shield frem the hot celi:
I. disconnecting instrument, wafer, gas, etc.,
lines to reactor and shield,

2. removing pump drive motors and adaptors,

3. removing control-rod drive assembly (but not
rod),

4. cutting Nak pipes just outside the water shield,

5. removing all but two of the capscrews that
retain each fuel and sodium pump body in its
pump barrel,

6. removing most of the bolts that aoitach the
reactor and shield support bridge to columns,

7. drcining and removing lower pertion of woter
shield,

8. cutting fuel drain lines above valves,

9. draining and removing balance of water shield,

10, attaching reactor support bridge to crane,

11. removing last bolts attaching reactor support
bridge to columns,

12, moving reactor and lead shield assembly to
hot cell and attaching to disassembly fixture,

13, cutting off NaK menifolds just above lead
shield and removing from cell,

14, removing Ne pump bodies and removing from

cell,

15, removing fuel pump bodies and removing from
cell,

16. removing ftop, bottom, and side sections of
lead shield,

17. removing ZrFAevcpor trap.

It is evident that care should be tcken in the
detailed design of the ottachment of the reactor
support bridge to the celumns to facilitate the
removal of the final attachments prior to lifting ouf
the reactor shield assembly. It is also important
that the water shield be designed to permit rapid
disassembly by the mechanics who must enter the
cell to remove capscrews, etc. Similarly, it is
important that the lead shield be designed so that
it can be disassembled readily, with remote han-
dling equipment, in the kot celi. This will probably
require special design features.

Throughout the disassembly of the reactor
pressure-shell assembly, dimensional data must
be token at frequent intervals. For examgple, the
first step after removal of the shield will be the
determination of the key dimensions of the pressure
shell. The over-all height will be measured from
the level of the pump mounting flanges fo toth the
north and south heads. A cut can then be rmade ot
the equator through both the pressure shell and the
pressure-shell liner. A cut will also be taken
through the thermal sleeves around the heat ex-
changer outlet tubes ot the south end, and the
sleeve that attaches the island assembly to the
notth head will be severed. This will permit the
removel of the island assembly and the southern
half of the pressure shell. The core shells,
pressure shell, ond heat exchanger can then be
examined., The most important of the dimensional
data will show the tendency of parts to warp,
shrink, or otherwise distort during the coriginal
welding and assembly process and during festing.
For example, distortion might cause an interference
between the impellers and the staticnary elements
of the pump, which, in turn, might cause mal-
functioning of the pump, Since changes in the
thickness of cooling onnuli might lead to poor
flow distribution and hot spots, dimensional data
on the cocling annuli for the core, refiector, and
pressure shells should also be recorded. This
will require measurement of the diameters of the
beryllium hemispheres and the various shells, and,
of course, it is presumed that a comparison with
dimensional checks will be made during inspection
prior to and during assembly.

The welds that attach the pressure-she:l dome
to the north head around the rocts of the fuel pump
barrels can be milled out, and cuts can ke made
through the transfer tubes that attach the reflector
assembly to the north head, as well os through the

93
thermal sleeves and heat exchanger outlet tubes
at the north head. The two regions can then be
separated for inspection. The reflector shell can
be cut out and the beryllium can be removed so
that bcth can be inspected for corresion, spacer
fretting, ond thermal cracking, After visval and
dimensional inspection, representative sections
of key elements of the various assemblies can be
cut out, ploced in lead pigs, and removed to smaller
hot cells for metallurgical examirnation, chemical
analysis of deposits, stc. Many areas should be
carefully dyeschecked for thermal cracks that
might have been induced during the operation.
When plated-out activity is evident, the relatively
short half life of the activity may make it desirable
to defer work on that part and to work on other less
vital but much less active parts having long hailf
lives.

The equipment in the hot cell must include strong
manipulators for handling segments of the reactor
assembly, heavy cutting equipment for sectioning
the reactor intc major subassemblies, and equip-
ment for cutting various samples from the sub-
assemblies. An Inconel cutting torch might be
useful for some of this work, However, the cutting
torch should not be used where there is a large
amount of octivity, becouse dispersal of the
activity into the atmosphere of the hot cell would
make decontamination of the cell after completion
ot the disassembly operations exceedingly diffi-
cult and would greatly aggravate the problem of
loading samples into lead pigs for removal from
the cell. In general, it seems that no torch cut
should be made where the material in the cut would
have a total activity in excess of 0.001 curie,
while the total amount of activity in all cuts con-
templated with the lnconel torch should not exceed
G.01 curie. Cuts withamilling cutter or an abrasive
saw should be mede in such a way that the ma-
terial removed by the wheel or cutter will either
be washed info a sump with ¢ stream of liguid
coolant or will be drawn inte o suitable dust
collector by a vacuum-cleaner type of system.

The air-borne contomination con probably be
reduced markedly by the use of saw and milling-
cutter type of equipment, which remove relatively
coarse chips, rather than cutting forches or high-
speed grinding or abrasive types of cutting-wheel
equipment, which yield a fine dust.- This means
that it may well prove worth whiie to adopt, as a
basic premise, the requirement that all cutting
operations be made with relatively low-speed sows

94

or milling cutters in zones in which the specific
activity is fairly high. An important zone from
this standpoint is the region around the pump
barrels in the north head. |t must be expected that
the specific activity of the material in the north
head will be cbout ten times as great as that of
material clcse to the equator. Hence much of the
total activity anticipated in chips formed during
the separation of the major subassemblies will be
in those formed during disassembly of the north
head.

Another measure that might be taken to reduce
air-borne contamination would be to fill the reacter
with a thin solution of strippable varnish. Before
removal of the reactor from the test cell, the varnish
could be drained and the reactor fluid circuits
could be dried with air that could be discharged
through the regular off-gas system. Such a step
may be essential to reduce the air-borne contami-
nation to acceptable limits for the period between
the severing of the line to the dump tank and the
closing of the hot cell,

It is quite evident that space inside the hot cell
will be at a premium, because the varicus sub-
assemblies of the reactor will take up much more
space than the assembly. In this connection it
should be noted that the greatest activify will be
in the beryllium, the core shells, and the variocus
parts that will be wetted by the fuel. If these parts
can be separated from the others and placed under
a portion of the lead shield, the activity of the
unshielded items in the cell will be much lower,
and it will be much easier to open the cell for
insertion or removal of lead pigs for sample re-
moval.

[t is evident that a few quite special handling
fixtures will be required in order to position the
reactor ond the various major components during
the disassembly operation. it will probably be
necessary, before cutting into the surfaces that
have been wetted by fuel, to close the hot cell,
since the air-borne activity from the reactor is
likely to be substantial. Similarly, it may be
desirable to coat the inside of the reactor dis-
assembly ceil with a plastic sheet to facilitate
decontamination after the disassembly operations.
A similar strippable coating on the major items of
equipment in the hot cell might also prove to be
advantageous.

It will be necessary to install a zinc bromide
window to permit observation of the disassembly
operations. It would seem desirable to design the

 

 
 

cell so that the major items of equipment used in
the disassembly operation can be remcved for
servicing. Possibly these large items may be
differentiated into two groups. The first group
would include oll ports that will actually come in
contact with the reactor surfaces which have
large amounts of activity, and the other group
would include the baiance of the mechanisms. The
first group might be considered to be expendable
and thus might be left in the cell. The remaining
equipment, which should include the heavy and
expensive mechanisms, might be designed so that
it could be removed foirly readily from the cell

and setviced outside. This will be possible if the
ait-borne contamination can be kept fo a total of
about 0.1 curie during the course of the disassembly
operation.

it is evident that once the hot cell is equipped it
could serve as a general-purpose hot cell for reuch
disassembly work on such items as in-pile loops,
and therefore it might be worth while to make this
hot cell as well suited to other types of work as
can conveniently be done. This would imply that
manipulator and cutting equipment might be de-
signed for more general-purpose operaticns than
necessarily required by the ART disassembly.

95
Part VI

STATUS OF DESIGN

AND DEVELOPMENT
 

STATUS OF DESIGN AND DEVELOPMENT

The following outline of the key design problems of the circulating-fuel reflector-
moderated reactor and of the present status of the problems serves as a summary of the
work that has been done and of the werk that remains to be dene to provide a sound basis
for the design of a full-scale aircraft power plant,

DEVELOPMENT PROBLEM
FUEL CHEMISTRY AND CORROSION

Corrosion

Harp tests and simple thermal-

convection loops

High-tempercture-differential, high-

velocity loops

Radiction Damage and Corrosion

In-pile capsule tests

In-pile loop tests

Physical Properties
Na F-KF-L.iF, NQF-BeF2, NaZrFs,

etc.
NaF-RbF-LiF

Other fuels and fuel carriers
Sotubility of UF4 and UF,
Methods of Preparation
Xenon Removal

Reprocessing Techniques

Na-to-NaK

Pressure losses for flattened-wire

tube-spacer arrangement

Heat transfer and endurance test

NaK-to-Air

Fabricability, performance, and
endurance tests {including study

of character of fcilure)

STATUS MAY 1956

Many favorable results

Many favorable results

Many favorable results

Many favorable results

Adequate data

Adequcte data

Adequate data
Adequate data
Adequate data
Some data

Adequate data

H!GH-PERFORMANCE HIGH-TEMPERATURE HEAT EXCHANGERS

Adequate data

A dequate data

More tests needed

REPORTS

ORNL-1515, -1609,
-1649, <1692

ORNL-1896, -1947,
-2012

ORNL-1896, -1947,
-2012, -2061

ORNL-1896, -1947,
-2012, <2061

ORNL CF-53-3-261

ORNL-2012

ORNL-1896, -1947,
~2012

ORNL-1896, -1947,
-2012

ORNL-1896, -1947,
-2012

99
DEVELOPMENT PROBLEM

Fluoridesto-NaK

Tube-to-header welding, endurance,

and performance tests

Effects of trace leaks and fabrica-

bility of spherical shell type

SHIELDING

Preliminary Designs

iLid Tank Tests of Basic Configura-

tions
Effects of thickness of reflector,
pressure shell, lead, and boron
layers
Estimated Fuil-Scale Shieid Weights
Effects of power, power density,
degree of division
Activation of Secondary Coolant
Estimated activation

Measurements for neutrons from

core
Measurements for neutrons froem

heat exchanger

Measurements of Short-Healf-Lived

Decay Gamma Rays

Experiments on Air Scattering

Effects of Shield Penetrations

STATIC PRYSICS

Multigroup Calculations

E ffects of core diameter, fuel-
region thickness, reflector thicke
ness, reflector peisons, and

special materials

Effects of end ducts

Critical Experiments

Critical mass with various fuel
regions — Na, fluoride, fluoride-

graphite

100

STATUS MAY 1956

More tests needed

Some date available

Many designs available

Adeguate data for preliminary

design

Adequate data for preliminary

design

Adequate data

Adequate data

Adeguate dats

Adequate data

Adequate data

Test in progress

Adequate data

Calculations in progress

Adeguate data

REPORTS

ORNL-1896, -1947,
-2012

ANP-53, Y-F1510,
ORNL-1575

ORNL-1616, -2012

ORNL-1575, -1947

ORNL-1575, -2012
ORNL-1616, -2012

ORNL-1947, -2012

ORNL-1896, -1947,
-2012, -172%

ORNL-1515, -1729,
C F"54‘7.5r

ORNL.-1515, -1896,
-2012

 
 

DEVELOPMENT PROBLEM

Control rod effects {rough)
End duct leakage

Danger coefficients for Pb, Bi,
Rb, Li’, Na, Ni, etc.

MODERATOR COOLING

Estimation of Heat Source Distri-

bution

Be-Na-lncone! Corrosion Tests
Harp tests

High-temperoture-differential, high-

velecity loop

Be diffusion inte Inconel and

embrittlement

Effects of temperature, AT,

surface-veliume ratic, etc.

Thermal Stress and Distortion Test

with High-Power Density

PUMPS

Shakedown of Pumps with Face-
Type Gas Seals

Model Tests of Full-Scale Pump

Endurance Tests of Full-Scale Pump

Fabricability of Full-Scale Pump

Impetler

POWER PLANT SYSTEM

Preliminary Designs

Performance and Weight Estimates

Effects of Temperature, Power

Density, Shield Division, etc.

STATUS MAY 1956

Adequate data
Test data needed

Adequate data

Good estimates completed

Adequate data

Some data

Further tests under way

Some data

Adequate data

Preliminary tests completed;

more required

Tests completed satis-

factorily
Tests under way

Tests completed satis-

factorily

Adequate data

Adequate data for pre-

liminary design

Adequate data for pre-
liminary design

REPORTS

ORNL-2012

ORNL-1517, 2061

ORNL-1692, -1816
ORNL-1692

ORNL-1864

ORNL-1771

ORNL-1947
ORNL-1947

ORNL-2012

ANP-57, ORNL-1255,
-1215, -1330,
«1509, -1515,
-1609, -1648

ANP-57, ORNL.-1255,
-1215, 1330,
-1509, -1515,
-1609, -1648

ORNL CF-54-2-185

101
DEVELOPMENT PROBLEM
REACTOR KINETICS

Theoreticel Analyses

ARE Temperature Coefficient

Measuremenis

Xenen Effects

HYDRODYNAMIC TESTS

Flow Separaticn at Core Inlet

Effects of Heat Generation in the

Bouvndary Layer

Magnitude and Effects of Well Tem-
perature Fluctuations Caoused by
Nonunifermities in Fuel Tempera=

ture Structure

FILL-AND-DRAIN SYSTEM
Preliminary Design

High-Temperature Tests

HAZARD ANALYSIS

FACILITY DESIGN

FACILITY CONSTRUCTION

REACTOR DESIGN

Detoiled Preliminary Layout

Detailed Drawings

Reactor Stress Anclysis

Creep-rupture properties of Inconel

under severe thermal cycling

Creep-buckling properties of
tnconel under severe thermel

eycling

102

STATUS MAY 1956

Preliminary analysis

completed

Tests completed satis=

fc:ciorifly

ARE and anclytical results

favorable

Many tests completed; work

continuing

Thecretical anclyses com-
pleted for ideal case; tests

in progress

Analyses and Tests in

progress

Design completed

Tests planned for latter part

of 1956
Completed

Completed

To be completed atter part
of 1956

Completed

To be completed summer

of 1956

Some data; tests under way

Some data; tests under way

REPORTS

ORNL CF-53-3-231,
ORNL-1835

ORNL-1816

ORNL-1816, -1924

Y'F]S‘] }g
ORNL-1692, -1947

ORNL-1701, -2061

ORNL-1835

ART Design Memo
Book

ART Pesign Memo
Book

ORNL-1947

ORNL-1947

 
DEVELOPMENT PROBLEM

Anclyticel and experimental stress

cnolysis

- Fabrication

Instrumentation and Control Re~

quirements
Test Program

Operating Procedure

DISASSEMBLY PROBLEMS

Activity of Major Components
Beryllium
Disassembly Procedure

Tool design

Hot cell design

STATUS MAY 1956

Work in progress

To be completed summer

of 1957

Delineation completed

Program being prepared

Procedure being mepared

Adequate data

Adeguate data

Work planned

Work planned

REPORTS

ART Design Memo
Book

ART Design Memo
Book

ORNI.-1864
ORNL.-1947

103
Appendix A

FLOW DIAGRAMS
 

&

 
 

FLOW DIAGRAMS

Fiow diagrams for the ART
appended in the following order:

1. Fue! Fill-and-Drain, Enriching, Sampling,
and Recovery Systems
Sodium System
Off-Gas System
Cell Pumps Hydraulic Drive Systems
Reactor Pumps Tube Oil System
. Main, Auxiliary, and Special NaK Systems
. NaK Pumps Tube Qil System
. Precess Air System
. Helium System

I0. Nifrogen Systfem

11. Compressed Air System

12. Process Water System

The following system of nomenclature has been
used in the flow diagrams, The designation of o
component consists of three letters followed by a
The first letter identifies the fluid
system; the second letter identifies the parficular
component in that system; and the third letter
identifies the power source with which the com-
ponent may be associated., Occasienally the
component is not associated with a particular
power source, and the third letter is omitted,
When there is more than one component with the
same set of letters, the letters are followed by a
number to differentiate between components,

The letter abbreviations are as follows:

systems are

numeral,

FIRST LETTER

Fiuid Systems
Fuel
Sodium
Ne K
Process air
Hydraulic fluid

Oil lubricotion

OXTP»XRZM

He lium

Process water

Compressed dry air

Nitrogen

Off-gas system (to absorbers)
Vent system (to ducts)

<XINE®

SECOND LETTER

Compenents
Blower
Qil catch basin
Hect barrier doors
Line
Heat economizer
Filter
Plug indicator
Air control fouvers
Motor
Pump
Radiator
Tank or sump
Package unit
Heat exchanger

XC=-VVTO—TMMrOUO®

THIRD LETTER

Power Source
TVA power A
B

Diesel power

Where appropriate the following designations have

been added:

Supply
Return
Breather
Hot

Cold
Upstream
Downstream

a.c 0o TTo Y on

107
 

      
     
 

 
  
 

JACKET COOLING WATER IN —
SEE DIAGRAM 12

 
   
 

WATER = HELIUM SUPPLY~ SEE DIAGRAM 9
SEE DIAGRAM {2

PCY, XF-13

  

XF-i4 JACKET-COOLING
©  WATERQUT—
SEE CIAGRAM 12

0 OFF-GAS SYSTEM -
SEE DIAGRAM 3

  

FILTER

    
 
  
   

PPy~
&

HYDRAULIC SUPPLY
SEE DIAGRAM 4

    

TO HYDRAULIC PUMP
DISCHARGE PRESSURE~
SEE DIAGRAM ¢

 
 
 
 

 
 
 
 

OlL SCAVENGE

 

          
       
  
 
 
 
  
  
 
   
  
   
 
 
 
    

LINE, He
sEE D) s JuuseoiLRETURN LUBE O1L RETURN
BREATHER BREATHER
LUBE OIL SUPPLY LUBE OIL SUFPLY

END AND CAP
N TUNNEL

   
    
   
  
 
  
 
  
     
    
 
 
 
       
  
  
  
 
 
   
   
 
  
   
  
  
   

TO BE REMOVED PRIOR TO
HIGH POWER OPERATION

FUEL OVERFLOW LINE

NITROGEN CODLANT FOR DUMP
WALVES, SEE CIAGRAM 10

VF=2

FiLL LINE FOR CARRIER AND
FOR INITIAL CONCENTRATE

 

A

 

SWAGELOK

    
  

 
 

SPECIAL NoK

SPECIAL NOK SYSTEM
B OUT "A"IN

SPECIAL NoK SPECIAL NaK SYSTEM
8" IN A'OUT

     
     
   
 

EN AND CAP
INCELL

 
 

OIL SCAVENGE LINE  He

 
 
 
   

 
  
   
   
  
  
   
     

AT 28 psig)
GF-16

sagaphET™
ORNL-LR-DWG 15049

 

HELIUM SUPPLY-
SEE DIAGRAM 9
TO HYDRAULIC PLMP DISCHARGE
PRESSURE — SEE DIAGRAM 4
SEE DIAGRAM 5
LEGENE

s CONTINUOUS SPEED INDICATOR
CSR CONTINUOUS SPEED RECORDER
HY HAND-ACTUATED CONTROL YALVE
HMA  HYDRAULICALLY DRIVEN MOTOR A,
HMB  HYDRAULICALLY DRIVEN MWOTOR 8
HS HAND~ACTUATED SWITCH

LA LEVEL ALARM

u LEVEL INDICATOR

HY MANUALLY OPERATED SLOCK VALVE
Pi PRESSURE INDICATOR

PT PRESSURE TRANSMITTER

FE FLOW ELEMENT

P PRESSURE CONTROLLER

€ POSITION -TRANSMITTING ELEMENT
n POSITION INDICATOR

CFi CONTINUOUS FLOW INDICATOR

CPR CONTINUOUS PREESURE RECORDER

ECY ELECTRICALLY ACTUATED CONTROL VALVE
HC HAND~AC TUATED CONTROLLER

= FUEL LINES

—===— NaK LINES

e GAS, AR, WATER, OiL., BISMUTH, OR VALUUM LINES
=== FLECTRICAL CONNECTION

—AoA PNEUMATIC LINES

€ SWAGELOK CAP ON END OF TUBE

 

F NITROGEN SUPPLY—
R ,,'_‘5‘ SEE DIAGRAM 10

 

P = LINE REDUCER

O —— LN mrEaK bEVICE

O — FLOW, iiters/24
: — TEMPERATURE , *F
— EQUIPMENT

E IDENTIFICATION

(@  PUSH BUTTON SWITCH

éd SOLENOID VALVE
PRESEURE-REQULATOR WALVE

T SWING CHECK VALVE

$<] THROTTLING VALVE
FRANGIBLE DISK VALYVE
{ NORMAL LY CLOSED)

[><] NORMALLY OPEX

o] NORMALLY CLOSED

 

 

FUEL RECOVERY TANKS {1250"F WHEN CHARG -
ING FIRST BATCH NITROGEN SUPPLY-
SEE DIAGRAM 10 .
PRE-POWER FUEL FILL-AND-DRAIN SYSTEM
FUEL RECOVERY SYSTEM J i SAMPLING SYSTEM ‘ l FUEL ENRICHING SYSTEM J ‘ AND REACTOR ‘ l POST-POWER SAMPLING SYSTEM ‘

Flow Diagram 1. Fuel Fill-and-Drain, Enriching, Sampling, and Recovery Systems,

 

 

109

 

 
—r

 

 

e a2 et T ™ R e e

 

— e

 

 
WATER RETURN {SEE DIAGRAM NO. 121

Y-
HV“WN 23

wY

 

   

HOCV X Wh-21{

&

 

  

 

SET POINTS

Hee

 

 

l
I
|
|
I
I
| .
!
I
I
|

l—neuuu sum.v JSEE DIAGAAM NC.9)
..........._._..,.,,_.___._..._.___.____I v
EOUIPMENT LpCATED QUTSIDE CELL |

T P

{2+min. TRANSIT TIME BETWEEN‘ SODIU"“‘ AND AUXILIARY EQUIPMENT PIT}

it S

v SET POINTS:
L $ 20 (P& ]
- 210

WATER SUPPLY ( SEE DIAGRAM NC. 12)

TO {ELL

  
   
      

*T“

/Hl ~FILLED HQUSING FOR CONTROL
ROD DRIVE MECHANISM

 

“WLSET AT 25 psig

 

  

 

TO CELL

 

 

TO TEMPORARY VACUUM PUMP
‘ !

" :
w-rafiy 20 &

 

    
 

T CELL VENT LINE
(SEE DIAGRAM NG, 3)

E : senomu:l
517

  

w32
OPENS AT 50 paig

 

 

(SET POINTS)

(SET POI TS5}

   

 

Ne EXPANSION TANX
(AUXILIARY)

EATED LINES

o~ nwrua: DisKs

 

HELIUM SUPPLY (SEE DIAGRAM uo 51

AND VENT LINE

 

TEMPORARY
HELIUM SUPPLY

  

 

N-21 HVAGN-t4
2

 

TEMPORARY vAGUI|

if'r:luc ,._,J.___

" - KEATED LINE

HELIUM SUPPLY (SEE DIAGRAM NO.2}

@—— — —-Yon-31
HVXGN-!Z

VN-3

LITHIUM
FILL LINE

 

No FILL LINE (HEATED) "~

W
w
GN-12

(; > /mr LINE FOR REMOVING No
' Nt

 

 

 

mfiiuuc SUPPLY (SEZ DIAGRAM NO. 8)

 

 

— i v o g o

 

 

 

 
 

NYDRAULIC SUPPLY {SEE DIAGRAM lo. 4.

(SEE O1AGRAM KO, 51

TO MYDRAULIC PUMP DISCHARGE PRESSURE —
(SEE DIAGRAM ND. 4]

Oll. SCAVENGE LINE

 

 

 

LUBE OIL RETURNg
BREATHER -

L -5
@ @ uAmTE’f"é CRuP L

 

 

 

LUSE QIL SUPPLY

Ne EXPANSION TANK

 

 

 

 

 

3 mgnetc For — @— _Eié

200-600

 

 

(STANDBY, 300 °F ; OPERATE, 550 °F
TOTAL VOLUME, 60 in? ; DIFFERENCE W

VOLUME BETWEEN UPPER PROBE LEVEL

AND LOWER PROBE LEVEL, 40 i} )

g3

ek
ORML- LR-DWG 18080

EQUIPMENT IDENYIFICATION

€S| CONTINUOUS SPEED INDICATOR
CSR CONTINUOUS SPEED RECORDER
CV  CHECK VALVE

ECV ELECTRICALLY ACTUATED CONTROL VALVE

LI LEVEL LIGHT
HCY HAND ACTUATED CONTROL VALVE
HY  MANUALLY OPERATED BLOCK VALVE
NPA SODIUM PUMP A
NPB SODIUM PUMP B
NXA SODIUM-TO-Nek HEAT EXCHANGER A
KXB SODIUM-TO~NoK HEAT EXCHANGER B
Pl PRESSURE INDICATOR
SV SOLENOID VALVE
PT  PRESSURE TRANSMITTER
CFI  CONTINUOUS FLOW INDICATOR
CLI CONTINUOUS LEVEL INDICATOR
PI CONTINUOUS PRESSURE INDICATOR
4 PRESSURE ALARM
| POSITION INDICATOR-~LIMIT LIGHTS
RELIEF VALVE
FLOW ELEMERT
FLOW ALARM
FLOW INDICATOR
. TEMPERATURE AL ARM
WATER FILTER
HAND -ACTUATED CONTROLLER
HAND ~ACTUATED SWITCH
TEMPERATURE -ACTUATED CONTROLLER
PY  PRESSURE ~-ACTVATED BLOCK VALVE

mM ANIO
P m <

o X
353

PRESSURE, prig

TEMPERATURE, *F

FLOW, 1it/24 hr

FLOW, gpm

 

 

 

 

 

HEATED LINE - PV
N-11

 

 

%]
| .|
B
153 1250
440 1050
287 1250
440 1050

% AUXILIARY NaK SYSTEM
{SEE DIAGRAM NO. 6)

 

 

"B W

 

 

 

 

MAIN NoK SYSTEMS
{SEE DIAGRAM NC.6}

3

a7

NOTE! SODIUM FLOW RATES ARE THOSE
ASSGCIATED WITH ONE SODIUM PUMP

MAIN NaK SYSTEMS
{SEE DIAGRAM NO. 6)

CELL

 

 

 

d T '
: DA
% \i“
nz
-
ws) B
1 1250

 

  
 
 
 
     

 
 

No DRAIN
LINE [HEATED}

{SEE DIAGRAM NO.1}

Flow Diagram 2. Sedium stg?m.

AUXILIARY NoK SYSTEM
(3EE DIAGRAM NO.6)

MAIN NaK SYSTEMS
(SEE DIAGRAM NO. 6)

MAIN Nok SYSTEMS
{SEE DIAGRAM NO.6)

FUEL FILL AND DRAIN LINES

 

 

 

o T

 

e —————— it S

TO HYDRAULIC PUMP DISCHARGE BRESSURE {SEE DIAGRAM NO. 3}
4= OIL SCAVENGE LiNE

LUBE OIL RETURN
BREATHER

LUBE OIL SUPPLY

(SEE DIAGRAM NO.S)

=

100>

5 txg 0

LINE BREAK DEVICE

BIMETAL TEMPERATURE ELEMENT
THERMOMETER

CAP WELDED OM LINE
THERMCGCOUPLE

SWAGELOK CAP ON TUBE
NORMALLY CPEN VALVE
NORMALLY CLOSED VALVE
THROTTLING VALVE

SOLENOID VALVE

1M

 
[ T ———— e ——e -

 

 

 

———— e ————— e ————

 

 

 
 

 

 

 

 

 

 

 

i-'ss

 

 

 

 

 

 

  
   
   

 

I
<

1 PI
T

L
i

332

 

 

 

 

 

 

 

 

 

 

'SEE PROCESS AlR FLOW
DIAGRAM 8

 

 

   

 

 

 

 

 

 

 

: I ‘{x-ss S :x-ia
' ) ‘ wm .-} ACTIVITY HS ACTIMITY -..___.- --..- ACTIVITY
. ) | Ll - - MONITOR MONITOR [ | (=103l [=07%chl. L _] MONITCR
' " l S x-ss X-57 {X—IG
‘ C e - | R @_ FILTER ‘ FILTER
FROM REACTOR FUEL PUMPS OIL cn‘rcu aAsms PR _ ] - :
o SEE DIAGRAM NO. 8~ 7 e . X-55 ‘ : o $xe1d
L Co . DOWNSTREAM SAMPLE F— - . DOWNSTREAM SAMPLE , <
e . L STATION —REACTOR - ‘ _ * STATION-REACTOR
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'SEE FUEL SYSTEMS' FLOW. l l a TV L e o REACTOR VENT SYSTEM -
. IAGRAM NO.V =
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sev | b Tav 1 o LAZY R
oy I L el CONTROL
‘ PE l ‘ ,_H A
FROM REACTOR FUEL R | I EED ple
PUMPS BREATHER LINES oy o - -
SEE DIAGRAM NO. 5 : l T ™~ CELL VENT SYSTEM
X-42 | ‘. Ry ‘ .
" %-44 E
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sCV | T __DISASTER IN CELL
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 SEE SODIUM SYSTEM -
FLOW DIAGRAM NO. 2 -

 

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70

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UPSTREAM SAMPLE p—emee———
STATION ~CELL

VENT SYSTEM ‘E"E
-
1 X-47, X-49.4

 

 

 

 

 

 

 

HV X-48
LAZY ROD
CONTROL

VENT HOUSE

Flow Diagrdm 3. Off-Gas System.

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—_— e CHARCPAL ADSORBER

IN CHARCOAL ADSORBERS THE DASHED LINES
REPRESENT EMPYTY PIFE AND HEAVY LINES
REPRESENT CHARCOAY FILLFD PIPE.

JO>MDMZ%XXX

       
   
 

ACTIVITY
MONITOR

iperTT.
ORNL-LR-DWG 1505!

LEGEND

SWING CHECK VALVE
HAND-ACTUATED SWITCH -
MULTIPOINT ACTIVITY RECORDER
MULTIPOINT TEMPERATURE INDICATOR
HAND-ACTUATED CONTROL
HAND-ACTUATED BLOCK VALVE
PRESSURE INDICATOR

PRESSURE TRANSMITTER
PRESSURE ~ACTUATED BLOCK VALVE
RADIATION ALARM
DISASTER~ACTUATED BLOCK VAWVE
VALVE OPEN

VALVE CLOSED

THROTTLING VALVE

SOLENOID VALVE
CHECK VALVE °

EQUIPMENT IDENTIFICATION °
PRESSURE (psig)} .
FLOW (ctm).

TEMPERATURE , *F
ELECTRICAL CIRCUIT

-

WATER

 
 

SEE WATER SYSTEM
DIAGRAM NO, {2

SEE. WATER SYSTEM
DIAGRAM NO. {2

113

 

 
 

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¢ FLUD PUMP SPEEC
INDICATOR {SEE DIAGRAM 1}

SNUBBER AT 7.8 peig

86,2 hp MaX
2400 cpm MAX

   

PUMP
HYDRAULIC MOTOR
HMA-1{

TO FPA (SEE DIAGRAM £)

      
  
  

XX

TO SODIUN PUMP
SPEED INDICATOR
(SEE DIAGRAM 2]

   
 
 
 
 
  

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SNUBBER

  

OPENS AT 7.8 psiq

30 hp MAX
3300 rpm MAX

  

  
   
   
    
  
 
  
  

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HYDRAULIC MOTOR
HMA-2
TO NPA (SEE DIAGRAN 2}

 

 
   
     

  

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"TANK DESIGNED FOR
210 peig MAX . .
PRAGLIC FLUD" WORKING PRESSURE, 135 paig

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FILL—AND-DRAIN LiNE

    
 
 

   
 
  
  
  
 

HYDRAULKC FLUHD
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INDICATOR
(SEE DIAGRAM 1)

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SPEED INGICATOR

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SNUBBER AT 7.5 psig

30 hp MAX
3300 rpm MAX

  

HV
H-45

PUNP
HYDRAULIC MOTOR
HME -
TO NPg (SEE DIAGRAM 2}

    
   
  

   

    

   
  

 
  

  
   
   

 

                
   
 

   
 

      

 

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ORNL-LR~0WG 13082

CONTINUOUS CURRENT INDICATOR
FLOW ALARM

FLOW INDICATOR

LEVEL INDICATOR

MULTIPOINT TEMPERATURE RECORDER
HAND =ACTUATED BLOCK VALVE
PRESSURE INDICATOR ’
PRESSURE TRANSMITTER
HAND-ACTUATED CONTROL VALVE
CONTINUOUS POSITION INDICATOR
TENPERATURE ALARN
TEMPERATURE - ACTUATED CONTROL VALVE
WOBBLE-PLATE POSITION SYNCHRO
POSITION SWITCH

ELECTRICAL SWITCH

POSITION INDICATOR

TEMPERATURE SWITCH

CONTROL STATION )
ADJUSTABLE POSITION SWITCH
LEVEL TRANSMITTER

SWING CHECK vauvE

TEMPERATURE WONCATOR

FLOW ELEMENT

PRESSURE, peig

PRESSURE, in. Hg

TEMPERATURE, 5F

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WASTE OIL " 10 PORTABLE fULL LOAD I SUPPLY LINE
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SET BY MANUFACTURER
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Vi3 JHY  HYY Me13 017 LHV o
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I

Flow Diagram 5.

 

Reactor Pumps Lube Oil System.

 

117
 

 

B R T

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JO-oul SYSTEM 30-gol SYSTEM

MOTOR BREAKER MOTOR BREAKER MOTOR SREAKEN

. TAWI?!R TACHOMETER' TACHOMETER

GENERATOR GENERATOR oL CATCH i GENERATOR
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PRE-POWER PRE-POWER
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NoK ACTIVITY
MONITOR

NaK ACTIVITY

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8% MiIN
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{SET_POINT
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18 13
12 gol)

TG A SEE DIAGRAM T
HELIUM SUPPLY

PURGE GAS VENT B PUMPS TO

INTERMITTENT) FiLL~

WATER AND COMPRESSED AIR RETURN TC ENTRAINMENT SEPARATOR —SEE DIAGRAM 12
WATER SUPPLY TO FILTERS — SEE DIAGRAM 12

 

COMPRESSED AIR SUPPLY TO PLUG INDICATORS AND COLD TRAPS —SEE DHAGRAM 1Y

AUXILIARY SYSTEMS ’——l———. MAIN SYSTEMS A

Flow Diagram 6. Main, Auxiliary, and Special NaK Systems,

AGRORET
CRNL~LR - DWS 130540

TO OR FROM REACTOR

119

 
 

 

r -

 

 
   

S|

        
 
   
    
  

90 MAX

89 MiN 129 Max

3 MIN

121 MAX
93 MIN

Ha MaAX
L

THIS CIRCUIT COOLS THE OUMP TANK JACKET

         
 
  
 
  
   
 
 

(CASHED LINES DENCTE MATERIAL

HEATE
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70 gt SYSTEM

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BREAKER

 
  

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GENERATOR GENERATOR

   

  

 
 

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NaK ACTiVITY

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TEMPORARY
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o e e e e o e o e k. S i et S — o —— — e ekl TR ot ot i o i S

 

15 1+
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MR TO PLUG
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; MAIN SYSTEMS A SPECIAL SYSTENS (IN MoK COOLING CHWAMBER)

—4—-

Flow Diagram 6 (continved)

P

 
    
 
  

ORNL+LR=-DWG 15034p

MAIN SYSTEWS 8

121

 
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TO OR FAROM REACTOR

TO gal SYSTEM

MOTOR BREAKER

Iléu‘ OMETER
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MONITON

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PLUG INDICATOR
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MoK FILYER "y
ORANY LINE K-83

TEMPORARY
FILLING
EQUIPMENT

s 143
12 gai)

VIATIRE RETURN TO DRAIN
COOLING WATER 10 PLUG INDICATORS AND FILTERS L)
COMPRESSED AIR TO PLUG n{mcm)fls AND FILTERS

|_-—— MAIN SYSTEMS 8 ]

 

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HV

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MONITOR

KLB-11(C)

70 gul SYSTEM

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FILTER
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TEMPORARY
FILLING
EQUIPMENT

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FILL AND

 

Flow Diagram 6 (continued)

 

VK- SCV

TO pucY
SEE DIAGRAM &

™ DUCT
SEE DIAGRAM 8

obecTET
ORNL-LR-DWG 15034¢

CFR -~ CONTINUOUS FLCW RECORDER
€5 ~ CONTROL STATION
SCV — SWING CHECK VALVE
Fl = FLOW INDICATOR
FE ~ FLOW ELEMENT
LUl = LEVEL INDICATOR
21= POSITION INDICATOR
MAN = MULTIPOINY ACTIVITY RECORDER
AFE - RESTRICTOR FLOW ELEMENT
HV ~ HAND-ACTUATED BLOCK VALYE
Pl - PRESSURE INDICATOR
CS! = CONTINUOUS SPEED INDICATOR
AWM ~ RECORDING WATTMETER
FA = FLOW ALARM
LA - LEVEL ALARM
LMD ~ LEVEL—MEASURING DEVICE
CLA = CONTINUOUS LEVEL RECURDER
PT ~ PRESSURE TRANSMITTER
HE ~ HAND-ACTUATEQ CONTROL
MFI—~ MULTIPOINT FLOW INDICATOR
M3 = HANO~ACTUATED SWITCH
HCV = HAND=ACTUATED CONTROL VALVE
CLI - CONTINUOUS LEVEL INDICATOR
EV ~ ELECTRICALLY ACTUATEC BLOCK VALVE
GF — HELIUM FILTER

] remeeratune, °F
—————w—  NaK LINES, NAJOR FLOW
—_—

NaK LINES, FILLING, FILTER, WDICATOR, ETC.
gt PREUMATIC LINE

e ELECTRICAL CONNECTION
PRESSURE, p8i9
SWAGELOK CAP ON END OF TUBE

SOLENDID VALVE
_NORMALLY OPEM VALVE
CONTROL VALVE
NORMALLY CLOSED VALVE

FLOW, gpm
FLOW, ctm AT GO0® F AND 30 in. Mg

FLOW, liters /24 hr

LINE REOUCER OR EXPANDER

z +OOO:xxz~T[>

CHECK VALVE

123

 

 
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120 max WP
93 MIN ORNL-LA-DWG 150484p

12t MAX
23 MiIN

{8 MaX
WIN

FROM PUMS

PUMP

WEATED {DASHED LINES DENOTE MATERMAL THIS CIRCUIT COOLS THE DUMP TANK JACKET

OTHER THAN NoK!

35 Mk - { TO gal SYSTEM

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‘ z

MOTOR BREAKER

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. GENERATOR

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TANK TANK
OCA-4 oce-4

PRE-POWER ! PRE-POWER
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25 n?
(184 gol} @-

'
AR TO PLUG INOICATORS AND FILTERS
WATER TO PLUG INDICATORS AND FILTERS
WATER RETURN TO DNAIN !

SPECIAL SYSTEMB {IN SPECIA| NeK COOLING CHAMBER) _-‘L—= MAIN SYSTEMS @

 

{
Flow Diagram 6 (continued) ;
] 121

 

 

 
 

 

 

 
 

MAIN DUCT

 

——— e —— SPECIAL EQUIFMENT DUCT

 

COMPRESSED AIR SUPPLY
{SEE DIAGRAM NG 11)

 

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oo e, l . ALL HEAT BARRIER DGORS HYDRAULICALLY ACTUATED
MoTOR
—_— - L = ALL LOVVERS uvom;uucnuv ACTURTED " ‘:E: :{?f am ':,‘:,‘,:
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82,300 oo 1E =% N
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DUCT-ANNULUS BLOWER ABA-3 Tt
=

 

 

 

 

 

 

 

 

 

 

 

 

 

Flow Diogram 8,

b o

. .,
‘ i 8 AR SAMPLE LINES TO NoK LEAK ~Ja ]l fxal
e i, . . e e i, e, il WA e DETECTOR AND ALARM o0 1T
4000 venr ¥ m o {70 BE DEFERMINED) ¢
l - ..
— O G
DUCT-ANNULUS COOLING AIR
| o-—OPPOSED-ACYION LOUVERS:
by
-« -
12,800 4 RADIATOR AIR BYFASS (SEE IAGRAM NG. 8)
NeK OUT  NaK IN
—p ) ————
e e e 2l [ _D: »_
MOTOR 1 - ] v —————] - :
& "‘ison. —— | ] i - «ra-a ] TRl i —
b 2|4 % LS IQ’ - 2
— [sPR] m E———— 9
E n’sm:m. BLOWER ABA-4 = o % @_ -
12,500 D: m
| | ! _[sPr] ——— [6zs]
- L] ] - LI =2
, Y9l s /= | v MTR I’ -
i 1 e AT ‘ KRB-4 ——x :
— - - — :
éb MOTOR i w14 _ _[skR] _=
30 hp -—X | L]
]
— SECIAL BLOWER ASB-4 L%— —————— L M MEAT BARRIER DOORS N RADIATORS "N HEAT BARRIER DOORS
el —_—————— jolrm (
Cr— 430
4000 i o t:r < @ RADIATOR AIR BYPASS B
o<
) ] TVENT recs
B——Q | & ) K
’ -
DUCT -ANNULUS COOLING AIR
L wol I &2\ F —™ : :
5h IS /
"1
it [~
rfim-muuws BLOWER ABA-5

Process Air System.

 

 

'

MYDRAULIC CONTAINERS VENT (SEE DIAGRAM NO. 4}

SO,
OANL LA = OWE 18090

 

CELL OFF-GAS
(SEE DIAGRAM NO.3)

 

REACTOR OFF -GAS
{SEE DIAGRAM NO.3)

2
O

3
o

 

 

NeK PUMPS LUBRICAT

REACTOR PUMPS LY ’

b

NG SYSTEM RESERVDIRS (SEE DIAGRAM NO. 7)

: TING SYSTEM SUMP TANK VENT (SEE DIAGRAM NO. 5]

 

MoK DRAIN TANKS th‘f {SEE DIAGRAM NO. 6}

NoK PUMPS YENT t‘J

£ DIAGRAM NO. 8)

 

WASTE LUBE OIL §

MP YANK VENT (SEE DIAGRAM NO. 31

 

 

 

EQUIPMENT 1DENTIFICATION

CLOSED-POSITION INDICATOR SWITCH
CONTINUOUS PRESSURE -DIFFERENTIAL NECORDER
CONTROL STATION

CONTINUOUS TEMPERATURE RECORDER
MULTIRCINT TEMPERATURE INDICATOR
MULTIPOINT TEMPERATURE RECONDER
OPEN-POSITION INDICATON SWITCH
PHESSURE-DIFFERENTIAL TRANSMITTER
PRESSURE INDICATON

SHUTTER POSITION INDICATOR

SHUTTER POSITION RECORDER

SPRCIAL SWITEH (CENTRIFUGALY
INTERMEDIATE - POSITION INDICATOR SWITCH
CLOSED LIMIT SWITCH

OPEN LiMiT SWITCH

NeKt PRESENCE IN AtR ALARW

NeK PRESENCE IN AIR INDICATOR

PRESSURE, in. H0

FLOW, ¢tm

TEMPERATURE, *F

ON-OFF LIGHTS ADJACENT 70 ITEM INDICATED
PANEL ~MOUNTED LIGHT TO INDICATE CLOSED SWITCH
COMPRESSED AIR LINES

THERMOCOUPLE

THREE ~WAY VALVE

SOLENOID VALVE

127

 

 

 
 

 

 

 

A e

 

C

 
 

SONPTITR—
ORNL -LR-0OWG 15058

 

 

 

 

 

 

 

 

 

 
 
 
   

 

 

 

 

 

 

 

 

 

5100 MAX_ 1100 MIN 2
G-34
:5?\'; HV X 6-63 ‘
HV HV
. ~$><} . TO REACTOR FUEL PUMPS
' ‘ ' ' - 6-60 PCV 6-62 (SEE DIAGRAM NO. 1)
- ‘ . {wve-32 = E '
v Hv X6-67 . - EQUIPMENT IDENTIFICATION
! 631 T TO REACTOR SODIUMIPUMPS
XD Hvxs-aa G-66 : (SEE DIAGRAM NO. 2 SCV SWING CHECK VALVE
- [ ASV 1000~2000 : 2 FE  FLOW ELEMENT
A ' FI FLOW INDICATOR
Y + - ' HY X G- 71 FM  FLOWMETER
. ‘ TO FUEL EXPANSION JANK LEVEL INDICATOR HS  HAND-ACTUATED SWITCH
6-70 { SEE DIAGRAM NO.11 HV  HAND~ACTUATED BLOCK VALVE
Lo '+ 500-1000 g v PA  PRESSURE ALARM
HC HAND-ACTUATED CONTROL
HY X G-75
A PCV PRESSURE CONTROL VALVE
TO FUEL OVERFLOW LENE LEVEL INDICATOR POT PRESSURE~DIFFERENTIAL TRANSMITTER
(SEE DIAGRAM NO.1)§ - - Pl PRESSURE INDICATOR
TO FUEL RECOVERY $YSTEM PS  PRESSURE SWITCH

 

(SEE FUEL FLOW DIAGRAM NO.1) RFE 'RESTRICTOR FLOW ELEMENT
' ‘ ASY AUTOMATIC SHUTOFF VALVE

 

TO POST POWER FUEL] SAMPLE SYSTEM
(SEE FUEL FLOWDIA RAM NO.1)

 

 

FLOW, liters /24 hr

 

TC FUEL-ENRICHING SYSTEM AND FILL-AND-DRAIN TANK
(SEE FUEL FLOW DIA ?RAM NO. 1}
) FLOW, cmf

TO CONTROL ROD SObIUM SYSTEM
{SEE DIAGRAM NO. 2

 

PRESSURE, psig

 

 

 

TO LITHIUM INJECTIQN TANK
(SEE SODIUM FLOW PIAGRAM NO. 2)

 

OPEN VALVE

 

 

 

 

 

 

CLOSED VALVE

 

 

® | [Pl o

 

 

 

TO NaK FILL-AND-DRAIN TANKS
({SEE NaK FLOW PIAGRAM NO.6)
3

THROTTLING VALVE

Y

 

4
‘ CHECK VALVE

771 x DO

    
  

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

| J; ' 10-12 REDUCER GR EXPANDER
e HVXG-42
HV HY L
G-22 v < ] —<} TO A NoK PUMPS
HVX6-23  6-25 JHV G-39 \ PCV G-41 : {SEE NaK FLOW [BIAGRAM NO. 6)
OPENS AT ~ 6-40
280 psig ‘ “
' (P 2200
250 [ |
A L 10-12
HV X 6-46
I P > O 8 NeK PUMPH
{SEE NaK FLOWDIAGRAM NO. 6)
I
1
— e 0 REACTOR PUMP LUBE OiL RESERVOIR
U ReALITUN FU
[ P1 | {SEE DIAGRAM NO. 5)
l L
I Hv X6-50 A
4 TO NoK LUBE OlL RESERVOIR
G-47 " PCV G-49 {SEE NoK LUBE OIL SYSTEM DIAGRAM NO. 7) -

VEHICULAR ENTRANCE : P G-48 AUXILIARY EQUIPMENT PIT

Flow Diagram 9. Helium System,

‘ 129

 

 

 
 

st i et 8 s e e

 

.C

i

 

 
1

   
    
    
 

EXISTING RELIEF VALVE
M-15"QcET POINT 1100 psig -

EXISTING RUPTURE DISK;_

EXISTING

  

EXISTING RELIEF VALVE;
SET POINT 4100 psig

   

  
  
 
    
    

EXISTING RUPTURE DISK; ©  mv
- SET POINT 1125 psig '
HY i

EXISTING

  

 

M-22 . M-24

EXISTING

  

EXISTING RELIEF VALVE;
SET POINT HOO psig

   
   
     

EXISTING

 
  
   
   

EXISTING RUPTURE DISK; - .
SET POINT 1125 psig HV
HY

b

    
 

M-29
EXISTING

M-27 .
HV

STATIONARY NITROGEN CYLINDERS,
WILL HOLD 10,500 scf AT 1000 psig

 

GAS TRAILER
30,000 scf
AT 1800 psig

 

 

 

 

GAS TRAILER,
30,000 scf
AT 1800 psig

 

 

 

 

 

EXISTING

 

 

[ P |
]
4000 psig

HY X M~-32

 

 

 

 

 

HY X M-42

 

i

© e i

S it

5

 

     
  
  
  

8 NOR
116 MAX

250
260

 

 

 

 

 

 

RELIEF PRESSURE
300 psig

 

VEHICULAR ENTRANCE

 

m .
T}

i M-43 4
7 :‘ HV JE-D

 

 

  

 

o ( EV
-, ‘DISASTER IN CELL

.",TO AUTOMATICALLY
" "CLOSE BLOCK VALVE

 

  

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 
   
  
 

 

 

 

HV

 

 

e SPARE
M-44

AUXILIARY EQUIPMENT ROOM

 

INSTRUMENTATION BULKHEPDS (6)

 

 

FUEL FIlL-AND - DRAIN
UMATIC SUPPORT (SEE FUEL
ISTEMS FLOW DIAGRAM NO.1)

i

ROGEN SUPPLY FOR
L INSTRUMENTATION

L’ROGEN SUPPLY FOR
sv sv FILLING CELL

 

 

 

L TO HYDRAULIC SYSTEM JUNCTION
PANEL BULKHEADS
(SEE DIAGRAM NO. 4)

TC HYDRAULIC SYSTEM CONTAINERS
(S5EE PUMP DRIVE HYDRAULIC
SYSTEM PROCESS FLOW

DIAGRAM NO. 4) )

TO MAIN, AUXILIARY, AND SPECIAL
‘NaK PUMP SUMP LEVEL INDICATORS
.{SEE Naok SYSTEMS FLOW DIAGRAM NO.6)

RADIATOR PIT

>0

bt b4

 

 

 

 

 

 

CQRELGEME.
ORNL~-LR~-DWG 15091

EQUIPMENT {DENTIFICATION

HVY HAND~ACTUATED BLOCK VALVE

Pl PRESSURE INDICATOR

RV RELIEF VALVE

PS PRESSURE SWITCH

PA PRESSURE ALARM

HC HAND-ACTUATED CONTROL

XV DISASTER-ACTUATED BLOCK VALVE

FE FLOW ELEMENT

FI  FLOW INDICATOR

SV SOLENOID VALVE

PDT PRESSURE-DIFFERENTIAL TRANSMITTER
SCV SWING CHECK VALVE

HMC. HAND CONTROLLER

PCY PRESSURE -ACTUATED CONTROL VALVE
HS HAND SWITCH

AF  FLOW RESTRICYOR

FLOW, cfm

PRESSURE, psig

~—-—= ELECTRICAL CIRCUIT

PRESSURE -RELIEF VALVE
SWING CHECK VALVE

OPEN VALVE

CLOSED VALVE

SOLENOID VALVE'

LINE REDUCER OR EXPANDER

131

 
 

 

 

 

e — e et et e e S e e

 

 

 
 

 

 

 

 

 

 

 

 

 

 

   

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

   

 

 

 

 

 

 

 

e s Sy

e T

b

 

———— WATER RETURN
{SEE DIAGRAM NO.12)

 

 

 

 

 

 

 

 

 

 

 

25 sctm

Flow Diagram 11, Compressed Air System,

 
 
 

 

 

UNCLASSIFIED
ORNL-LR-DWG $5092

EQUIPMENT IDENTIFICATION

HCV HAND-ACTUATED CONTROL VALVE
HV HAND-ACTUATED 8SLOCK VALVE
PC PRESSURE CONTROLLER'

SCV SWING CHECK VALVE
HC HAND~ACTUATED CONTROL VALVE
Pl PRESSURE INDICATOR

PA PRESSURE ALARM
PCVY PRESSURE CONTROL VALVE

PS PRESSURE SWITCH
LCY LEVEL-ACTUATED CONTROL VALVE

LC LEVEL CONTROLLER
RF  FLOW RESTRICTOR

FLOW, scfm

PRESSURE, psig

VALVE OPEN

NTATION - SHOP

VALVE CLOSED

THROTTLING VALVE

CHECK VALVE

FOUR-WAY VALVE

- WATER SUPPLY {SEE DIAGRAM NO.12) .‘, <3 .
c-12 S
AIR [ P | ' CHEMISTRY LABORATORY
- -
1 5 HV
c-51 ! - o
. 5. SOLID STATE'LABORATORY
By b c-13 -
CPA i
AlR Hv HC LB
COMPRESSOR K LIMIT TO - N
: RF
, 8 sctm ‘
S
o c-14 c-15|F =16 HVAC-18
m uv¥c-s2 -~ YO PROCESS AIR BACKFLOW DAMPERS
‘ HV PCV THV (SEE DIAGRAM NO, 8)
: c7 b
“ HC
-Hv¥wc-52 g
0 I C-19 c-20| C-2t SR
Y pwe- - 4 2 TO MAIN, AUXILIARY AND SPECIAL NaK PUMP SUMP LEVEL INDICATORS
h 53 4 HY Y (SHE FLOW DIAGRAM NO. 6)
———— c-22 e
7we- FEL e ol MIST o >4
scv .
54 ”]I I TR Ny HV
: b TO AUTOMATICALLY CLOSE A
HCV I we-51. y 3 S ON "8" POWER FAILURE fi
HV XWC-50 % L
. 2. .’.‘A“’..ng"\:;':':_‘:;m;:“ — e TO NaK PLYG INDICATORS AND COLD TRAPS
i ey 2 200 (SEE NoK BYSTEMS FLOW DIAGRAM NO. 6)
i ] 20-30
e | ——— b Ev :j<: iy
o} i c-10 c-t c-24 |
¥ . PEW POINT INSTRUMENTATION - AUXILIARY EQUIPMENT PIT
A c-65 | ~20°F) '
~ -
r 11 [P ] .
! [ |
c-61 ! INSTRUME NTATION - AUXILIARY EQUIPMENT PANEL
PCV PS A
- AIR cP8 ' [ ] 15-20
AIR
l COMPRESSOR o
REACTIVATION - c-26 : !
AIR SUPPLY INSTRUMENTATION - AMPLIFIER RACK
fl HvXCc-~62 t HV { -l
: 15-20
v <G
i c-27
HY X WC-62 .| AIR COOLER AND : INSTRUME }rrmon- CONTROL ROOM
| waTER SEPARATOR oV
HE] *
wC-Y | ;
HY Lev OIL MIST : ‘
63 A b= z FILTER INSTRUMENTATION - TUNNEL
HevYwe - 61 < 100 i
scv /wc-64 Y > H
nv¥we-60 K
il
| INSTRUME

133

 
 

 

 

PR —— e " - -
— s . ————— s A et R et e B L e

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e —— e e —_

et e

 
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A W OVERFLOW & ‘ ToPT LEAD SHIELD Y
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o . w-ao¥uv-  w-stYHv  w-53XHV  wW-55XHv :
EXTERIOR o
INSTALLATIONS s o
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{ WHEN SWITGH }
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2 a4 e ’ <
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' . . ) ! X "‘ L [/, ' 2
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‘ . . : w-92 { BOTTOM) £
s AIR CONDITIONER . ‘ ;
s e 1 r _
LUINCH ROOM ; TO STORM DRAIN i @- | HY
e e EXISTING AIR COMPRESSOR - 10 CONTROL ROD Y \ W-87
; ' o L R gomum VAPOR CON~ Y L ‘ MTR
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' S TO FILL AND DRAIN ——8h— SEE DIAGRAM 1 7o
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. ' o SEE DIAGRAM
A N A | ‘ , L_9 -
SERVIGE SINK = : J} "
AIR CONDITIONER S — \ ‘ ‘@J__——fia—q-
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TO STORM DRAIN [ za | E T
HY L 4 NOT 1N USE oo h
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= CHEMISTRY LABORATORY - - " e
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SAFETY SHOWER | L L [ L
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HV F
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WY L NOTE: DISASTER N CELL - . DRAN TANK
w-58 o TO AUTOMATICALLY CLOSE
STRAINER l e T BLOCK YALVES. . I~ LEAD SHIELD
w-3 Y ‘
‘ Scv / W-25 | h | TO STORM i
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FROM X~10 TO CELL 2 o i i
RESERVOIR ANNULUS e TUNNEL :
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| Hv XwW—ig ‘ | - =
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TO CHARCOAL ADSORBER _ /' :
OR HOLDING POND W-23 AND STORM ORAIN ~ o
SEE DIAGRAM 3 . t‘ b
| R !
. .[ !\ .
\ PENTHOUSE ‘ ‘ VENT HOUSE ‘ k AUXILIARY EQUIPMENT AREA ‘ l CELL ‘

 

 

 

 

S
Flow Diagram 12, Process Water

 

 

System.

 

 

NN
ORNL~LR-DWG 15093

w-98
TO CELL DRAIN
LSET POINT
100 psig)

RUPTURE DISK

 

LEGEND —m——————

SCV  SWING CHECK VALVE

£V ELECTRICALLY ACTUATED BLOCK VALVE
FE FLOW ELEMENT -

Fl FLOW INDICATOR

HCY  MAND-ACTUATED CONTROL VALVE

HY HAND-ACTUATED BLOCK VALVE

MT}  MULTIPOINT TEMPERATURE INDICATOR
MTR  MULTIPOINT TEMPERATURE RECORDER
eA PRESSURE ALARM

Pl PRESSURE INDICATOR

ZA POSITION ALARM

s POSITION SWITCH

LCV  LEVEL-ACTUATED CONTROL VALVE

MAR  MULTIPCINT ACTIVITY RECORDER

PX PRESSURE SWITCH

TCV  TEMPERATURE-ACTUATED CONTROL VALVE
TE TEMPERATURE ELEMENT

XV DISASTER-ACTUATED CONTROL, VALVE

FLOW, gpm
EQUMPMENT IDENTIFICATION

==~ THERMOCOUPLE IN WELL

TEMPERATURE , *F

PRESSURE, psig

@DU

ORIFICE

(& TEMPERATURE-SENSING BULB

" REDUCER
NORMALLY OPEN VALVE
THROTTLING VALVE
NORMALLY CLOSED VALVE

FIXXY

RELIEF VALVE

135

  

 
———— | —em ¢ r———

 
Appendix B

PHOTOGRAPHS OF REACTOR MODELS AND STAGES IN THE

CONSTRUCTION OF THE ART FACILITY

........
HECRET
| PHOTOC 23368 i

 

 

B-1. Preliminary Model of ART Recctor, Heat Exchanger, and Pressure Shell Assembly.

 

139
 

 

B-2. Secticn Through the Core, Reflector, and Island of the ART Model.

 

140
171

 

   

F . oo

 INSTRUMENT
BULKHEAD

SUPPORT|
COLUMN

| FLOOR
| GRATING

B-3. Model of ART, Including NaK Piping and Cell.

| ~srcREE [
§ PHOTO 25608 8

 

 
142

EACTOR §

R

SHIELD

ATER

\u*'

NSTRUMENT
HA

]

S

MEBER

C

—
-
O
G
o
=
9y

N
5

LU

co

 

ith Shield Cut Away to Show Reactor.

Model of ART w

.

B-4

 

 
evl

 

B-5. Plastic Model of ART North Head.

~SEEREP
PHOTO 26473

 

 
7l

 

B-6.

Plastic Model of Nerth Head Showing L.ead Mocked Up in Wood.

SRR
PHOTO 26474 |

 

 
 

: "UNCLASSIFIED
o o " PHOTO 15410

 

B-7. View looking north showing the rear wall of the original Building 7503 and the early stages of excavation

for the extension tc accommodate the ART. In the foreground is the deep excavation for the reattor cell. Behind it

 

can be seen the excavation for the instrument and control tunnel. (Confidential with caption)

145
ori

 

UNCLASSIFIED
PHOTD 1581

 

B-8, View locking south from the original Building 7503 showing early steges of excavation and placement of concrete forms. On the right can be

seen the forms for the charcoal adsorber tank. The columns and beam on the left will support the low-bay portion of the building extension.

 
 

UNCLASSIFIED
PHOTO 15694

1]

:
:

¥
=]
ey
b
e
™.
-1
.
¥ o

 

B-9. Same as in Fig. B-7, 43 days later. In the center can be seen the reinforcing steel for the reactor cell support pad; behind is the form for the

instrument and control tunnel. To the right can be seen forms for the specirometer room and the columns and beams to support @ low-bay extension. In

Lyt

the bottom center is shown the reinforcing steel extending up from the charceal adscrber tank, {(Confidential with caption)
8Pl

3

UNCL ASSIFIED
PHOT O 15920

 

B-10, View looking south showing the reinforcing concrete footings for the stack and work in progress on placement of support columns for the high-

bay extension. In the right center is the charcoa! adsorber tank, and in the axtreme left center is the form work for the spectrometer tunnel,

 

 
   

e SO & |
ENCLASSIFIED
PHOTO 15758

 

 

 

B.11. View looking north showing the completed reactor cell support pad and the instrument and control tunnel. At the top center is shown construce

6¥1

tion in progress on the switchhouse. The form work for the top haif of the charceal adsorber tank is shown in the lower left. (Confidential with caption)
0sl

 

UNCLASSIFIED
PHOTO 16060

 

B-12. View looking south toward the stack foundation. In the foreground is shown support columns for the high bay and form work for the radiator pit

wall. To the right of the columns can be seen excavation for the blower house.

 
 

- UNCLASSIFIED
PHOTO 16058

 

 

B-13. View locking scutheast showing the spectrometer room at the lower left and the spectrometer tunnel in the

center, both of which are loceted under the low-bay extension to the building.

 

 

151
Zsi

 

UNCLASSIFIED
PHCTOC 16169

 

B.14. View looking north showing the steel framing for the building extension., In the fereground can be seen the stack base and the top of the char-

coal adsorber tank. The switchhouse can be seen at the left.

 
eat

 

 

B-15. View looking northeast showing the early construction
the main air duct for the primary cooling system.

UNCLASSIFIED
PHOTD 16280

&
&
:
5

work on the blower house. The opening in the main building at the right will recaive

 
CLASSIFIED
070 16317 K

 

B' 16-

View lcoking north taken from the south end of the high-bay extension and showing an initial stage in the
eraction cof the reacior cell water tank. The curved stee! ponels centairing nozzles, in the upper left, will be joined

into a ring and located on the anchor bolts shown ot the bottom center and thus become the support ring for the inner
vessel of the reactor call. {Confidentia! with caption)

154

 

 
 

 

ICLASSIFIED ]
PHOTO 16456 §

 

 

 

 

 

 

 

B-17. View locking southeast showing three of the five spectrometer tubes which provide a beam path between

the reactor cell and the spectrometer tunnel, which is underground beneath the lew bay. {Confidential with caption)

155
951

 

PHOTO 17152

 

 

B.18. View looking southeast across the construction work on the reactor cell, The form work, right center, is for the special equipment room walls.

The fiil shown in the top center is over the spectrometer tunnel and spectrometer tubes. (Confidential with caption)

 
 

LASSIFIED

L

UMC
PHOTO 17213

 

 

 

 

 

 

B-19. View looking southeast across the reactor cell showing a step in the placement of the top hemispherical head on the inner vessel of the cell.

 

The outer tank shown is the 30-ft-dia water tank. (Confidential with caption)

 

157
gs1

 

 

0 176

 

 

B-20. View looking north showing the facility in a late stage of the building alteration and addition program. In the right center is the 10sft-dic
75-fi-high stack for discharge of the air to be supplied by blowers to be located in the blower house at the bottom center. At the leftis thediesel-

generator house, which will accommeodote the five auxiliary units required for ART operotion.

 
   

LRI

L= UNCL ASSIFIED
w- PHETO 18233

 

 

B-21. View taken inside the blower house showing the supply end of the main cooling air duct and installation work on two of the four 82,000-cfm
[ ) 1 .t o I £ =1 . 9 - ol 1T AAM £ st E e 1 01 1 . L N1 4 1 o -1 ' o8 1 . 2 .1 .
DIOWeEerS.e vn LOTN 3 I108s Or TRe main qucy are me jv,vuv=crm pDIowers, WRICN Wil SUppIly QIr imro me GNnulius petwesn e Dulia g cencrerg ana e in=

sulated steel duct for the purpose of preventing overheating of the building structure. At the lower left can be seen the opening for the ramp which leads

661

down to the radiator pit beneath the ceoling air duct.
o9t

 

 

 

 

 

 

B-22, View looking west and from inside the cooling air duct showing the stainless steel facing over the insulated steel liner of the duct, The
supply end of the duct is the opening shown in the center of the photograph.

 

 
{91

 

will

8‘230

is the

 

View taken

 

NCL ASSIFIED
Q% 18153

 

inside of the main air duct from the base of the disch rge stack sh wing the porti n of the duct in which the NaK-to-air radiators

At the right ¢ nter is the ¢t facing of the cell water tank and the through the NaK pipe w ] Hass the duct

the tis the pit.
Zol

UNCL ASSIFIED
PHOTO 17733

 

B-24. View looking south across the reactor cell. In the foreground can be seen the top of the cell inner tank. At the top right is the concrete pen -
house which will house the primary coolant pumps and their drive motors. To the left of the penthouse are the roof plugs which will cover the special

equipment room immediately below. {(Confidential with caption)

 

 
£91

 

 

 

UNCL ASSIFIED
PHOTO 17734

 

 

 

 

 

 

B-25, View showing the southwest corner of the auxiliary equipment room. This roem, which was used during ARE operation as the heat exchanger

pit, is being modified to serve as the auxiliary equipment room and will house the lubricating cil pumping system and the hydraulic drive equipment,
y9l

n
STy
2
iy 2
e

=
gl
Pt

TUNCL ASSIFIED
PHOTO 17966

 

 

ottt it
IR M

Pyttt

 

 

 

 

 

 

 

 

 

B-26. This view, which was tcken in the switchhouse, shows the primary switchgear and instruments associated with the receipt and distribution of

the two power supplies to the facility. The switchgear and instruments on the left will serve the incoming purchased power from TVA, Those on the

right are for the locally generated power.

 

 
991

 

L UUNCLASSIFIED
. PHOTO 17960

 

 

 

 

B-27. View looking south from the north end of the original 7503 Building showing the top of the cell water tank in place as it will be located during

ART operation. This photograph was taken during the vacuum testing operation on the inner tank. {Confidential with caption)

 
291

UNCLASSIFIE
. PHOTO 17961

g

 

B-28. View looking north across the cell while the water tank top is in place. At the lower left can be seen the penthouse structure and the area in

which the main coolant pumps and motors will be located. (Confidential with caption)

 

 
L91

 

UNCLASSIFIED
PHOTO 17951

»

¥

 

B-29, View

e opened cell. The rectangular plates on the floor structure of the inner vessel will support the reactor. The
circular nlates will cunport the chislded fusl fillcandadrain tank ond fuslaracovery tank. The 24sin.edin nozzlae throush the side woll of the veczal will
provide for the penetration of the instrumentation and control lines, lube oii lines, hydraulic drive lines, cooling water lines
to the cell equipment. (Confidential with caption)

, gas lines, and heater leads

 
891

 

 

 

 

 

B-30. View looking north across the completed and painted cell. The floor structure was removed from the inner vessel at the time of the photograph
and therefore a good view is presented of the fluid distribution weirs located in the bottom of the vessel. The scalloped ring cutside the weirs is the
support for the floor structure. At the extreme bottom of the photograph can be seen the NaK piping sleeves with their expansion joints which penetrate
the two cell tanks. To the right of these sleeves can be seen portions of three of the 24-in.~dia spectrometer tubes. (Confidential with caption)

 

 
   

UNCLASSIFIED
PHOTO 18032

 

 

B-31, View locking southwest toward the modified bhuilding. In the foreground are the two top heads for the cell tanks which have been removed to

691

storage ouiside the facility. {Confidential with coption)
0LL

 

UNCLASSIFIED
'PHOTO 18033

 

 

Be32. View looking southwest toeward the diesel-generator house. In the lower left is the substation for the 13.8-kv purchased power supply from the
TVA system. The gas cylinders in the jower center of the photograph will be used for storage of nitrogen during ART operation.

 
tZ1

 

 

B-33, View lcoking west

iary power system.

inside

 

UNCLASSIFIED-
0 18154

 

the diesel-generator house showing a stage of the instaliation work on the five diesel-generator units for the auxil-
 

 
BIBLIOGRAPHY
 

 
 

BIBLIOGRAPHY
(Publications listed chronologically by subject)

GENERAL DESIGN

Nuclear-Powered Flight, Lex-P-1 (Sept. 30, 1948).

C. B. Eilis, The Atomic-Powered Aircraft January, 1950, ORNL-684 {April 26, 1950}.

Report of the Technical Advisory Board, ANP-52 (Aug. 4, 1950).

A. P. Fraas, Effects of Major Parameters on the Performance of Turbojet Engines,
ANP-57 (Jan. 24, 1951).

C. B. Ellis, The Technical Proh;’gms of Aircraft Reactors, ANP-64, Part | (June 8,
1951), Part 1 {Jan. 2, 1952},

G. F. Wislicenus, Hya’roa’ynamzcs of Homogeneous Reactors, ORNL CF-52-4-191 (April
30, 1952},

W. B. Cottrell (ed.), Reactor Program of the Aircraft Nuclear Propulsion Project, ORNL-
1234 {(June 2, 1952),

A. P. Fraas, Three Reactor Heat Exchanger—Shield Arrangements for Use with Fused
Fluoride Circulating Fuel, ORNL Y-F15-10 {June 30, 1952). i~

A. P, Fraos and M. E. LaVerne, Heat Exchanger Design Charts, ORNL-1330 (Dec. 7,
1952).

A. P. Fraas, Compenent Tests Recommended to Provide Sound Bases for Design of
Fluoride Fuel Reactors for Tactical Aircraft, ORNL. CF-53-1-148 (Jan. 13, 1953).

W. S. Farmer, Minimum Weight Analysis for an Air Radiator, ORNL CF-53-1-111 (Jan. 31,
1953).

R. C. Briant, Objective and Status of ORNL-ANP Program Presentation USAF Advisory
Committee, Washington, January 12, 1953, ORNL CF-53-2-126 (Feb. 13, 1953).

A. P. Fraas and C. B. Mills, A Reflector-Moderated ("zrculatz:zg +Fuel Reactor for an
Aircraft Power Plant, ORNL CF-53-3-210 (March 27, 1953}, .

W. S. Farmer et al., Preliminary Design and Performance Studies of Sodium-to-Air
Radiators, 0RNL-]509 (Aug. 26, 1953).

B. Lubarsky and B. L. Greenstreet, Thermodynamic and Heat Transfer Analyses of the
Aircraft Reactor Experiment, ORNL-1535 (Aug. 27, 1953).

R. C. Brian, Proposed ANP Program, ORNL CF-53-12-15 (Dec. 3, 1953).

R. W. Bussard ez al., The Moa’emtzng Coolmg System for the Reflector-Moderated
Reactor, ORNL-1517 (Jan 22, 1954).

W. H. Jordan, Proposed ANP Program — Ada’endum, ORNL CF-54-3-119 (March 24, 1954).

A. P. Fraas and B. M. Wilner, Effects of Reacror Deszgn Conditions on Aircraft Gross
Weight, ORNL CF-54-2-185 (May 21, 1954). [

F. A, Field, Temperature Gradient and Thermal Stresses in Heat-Generating Bodies,
ORNL CF-54-5-196 (May 21, 1954).

A. H. Fox et al., A Reactor Design Parameter Study, ORNL CF-54-6-201 (June 25,
1954).

A. P, )Frcas and A. W. Savolainen, ORNL Aircraft Nuclear Power Plant Designs, ORNL-
1721 (Dec. 3, 1954),

W. B. Cottrell et al., Aircraft Reactor Test Hazards Summary Report, ORNL-1835 (Jan.
19, 1955).

H. C. Hopkins, Vertical Components of Fuel Forces on Reflector and Island, ORNL
CF-55-2-142 (March 2, 1955).

C. S. Burinette, Fission-Product Heating in the Off-Gas System of the ART, ORNL
CF-55-3-191 (March 28, 1955).

A. S. Thompson, Empirical Correlation for Fatigue Stresses, ORNL CF-55-4-34 (April 5,
1955).

175
176

V. J. Kelleghan, High-Temperature Valve Information Summary, ORNL CF-55-4-83
(April 5, 1955),

A. S. Thompson, Allowable Gperating Conditions, ORNL CF-55-4-44 (April 11, 1955).

A. S. Thompson, Flexible Mounting Systems, ORNL CF-55-4-124 (April 11, 1955),

E. S. Bettis, ART Design Data, ORNL CF-55-4-87 {April 18, 1955).

W. B. Cottrell, The ART Gff-Gas System, ORNL CF-55-4-116 (April 21, 1955).

A. S. Thompson, Thermal Stresses in Tube-Header Joints, ORNL CF-55-4-15% {April 25,
1955).

T. J. Balles, Surface-Volume Ratios for Five Different Fluoride Fuel Systems, ORNL
CF-55-5-93 {May 12, 1955).

T. J. Balles, Temperatures and Pressures Resulting from Chemical Reaction Within
the ART Reactor Cell, ORNL CF-55-5-135 {May 17, 1955).

R. |. Gray end M. M. Ycorosh, Heat Transfer Design Data for ART Intermediate Heat
Exchanger, ORNL CF-55-5-120 {May 20, 1955).

R. . Gray and M, M. Yarosh, ART Main Heat Exchanger Designs, ORNL CF-55-5-141
(May 23, 1955).

B. L. Greenstreet and A. $. Thompson, Permissible Rate of Temperature Rise of Shells,
ORNL CF-55-5-175 (May 26, 1955).

A. S. Thompson, Flexible Support of Massive Bodies, ORNL CF-55-7-101 (July 19,
1955).

L. B. Andersen, Adsarption Holdup of Radioactive Krypton and Xenon, ORNL CF-
55-8-103 (Aug. 16, 1955).

H. C. Hopkins, Estimate of Nuclear Energy Generated During ART Accident, ORNL
CF-55-8-185 {Aug. 24, 1955),

L. E. Anderson, Symmetrically Loaded Cylindrical Skell with Fixed Ends, ORNL CF-
55-9-37 (Sept. 7, 1955).

J. M. Eastmen, Nuclear Aircraft Powerplant Controf, ORNL CF-55-9-121 (Sept. 23,

1955).

R. L.)MaxweH, Analysis of Various Designs for Intermediate Heat Exchanger Header,
ORNL CF-55-10-16 {Oct. 3, 1955},

W. J. Price, Radiation Levels Directly from Off-Gas Stack, ORNL CF-55-10-74 (Oct. 18,
1955).

W. J. Price, Neutron and Gamma Flux Distributions in ART, ORNL CF-55-11-7 (Oct. 18,
1955]).

M. H.)Cooper, Pressure Drop of Heat Exchanger Tube Spacers, ORNL CF-55-11-180
(Neov, 28, 1955),

C. W. Dollins, Lead Shielding for ETU/ART, ORNL CF-55-12-102 (Dec. 19, 1955).

W. J. Price, Preliminary Estimate of Activation of Various ART Materials by Thermal
Neutrons, ORNL CF-56-1-191 (Jan. 12, 1956).

W. L. Scott, Jr., Dimensional Data for ART, ORNL CF-56-1-186 (March 13, 1956).

D. L. Platus, Thermal Stress Analysis for a Proposed South End Configuration in the
ART, ORNL CF-56-5-166 {May 25, 1956},

F. A. Field, Rectangular Cross Section Cantilever Beams in the Plastic Range, CRNL
CF-56-6-85 (June 20, 1956).

COMPONENT DESIGN AND TESTING

G. H. Cohen, A. P, Fraas, ond M. E. LeVerne, Heat Transfer and Fressure Loss in
Tube Bundles for High Performance Heat Exchangers and Fuel Elements, ORNL-1215
(Aug. 12, 1952),

R. E. Ball, Investigation of the Fluid Flow Pattern in a Model of the *'Fireball’* Re-
actor, ORNL Y-F15-11 (Sept. 4, 1952}.

H. R. Jchnson, Valve and Pump Packings for High Temperature Fluoride Mixtures,
ORNL Y-F17-28 (Sept. 15, 1952).

 

 
W. B. Cottrell, Components of Fluoride Systems, ORNL CF-53-1-276 (Jan. 27, 1953).

W. B. Cottrell and L. A. Mann, Sodium Plumbking, ORNL CF-53-8-49 {Aug. 14, 1953).

H. J. Stumpf, Design Data and Proposed Test Schedule for Sodium-to-Air Radiators,
ORNL CF-53-9-102 (Sept. 8, 1953).

R. W. Bussard, Flat Plate Heat Exchangers for Reactor System Use, ORNL CF-53-
10-208 (Oct. 26, 1953).

B. M. Wilner and H. J. Stumpf, Intermediate Heat Exchanger Test Results, ORNL
CF-54-1-155 (Jan. 29, 1954).

H. J. Stumpf, R. E. MacPherson, and J. G. Gallagher, Test Resuits and Design Com-
parisons for Liguid Metal-to-Air Radiators, ORNL CF-54-7-187 (July 19, 1954).

J. E. Ahern, High-Conductivity Fin Fest Results, ORNL CF-54-8-200 (Aug. 27, 1954).

E. S. Farris, Summary of High Temperature, Liquid Metal, Fused Salt Pump Development
Work in the ORNL-ANP Project for the Period July 1950—]anuary 1954, ORNL CF-
54-8-234 (Aug. 1954).

R. W. Bussard and R. E. MacPherson, Thermal Stresses in Beryllium — Test No, 1,
ORNL CF-54-10-106 (Oct. 25, 1954),

D. F. Salmon, Turbulent Heat Transfer from a Molten Fluoride Salt Mixture to Sodium
Potassium Alloy in a Double-Tube Heat Exchanger, ORNL-1716 (Nov. 1954).

R. |. Gray, Temperature-Time History and Tube Stress Study of the Intermediate Heat
Exchanger Test, ORNL CF-54-11-69 (Nov. 30, 1954).

L. A. Mann, ART Reactor Accidents Hazard Tests, ORNL CF-55-2-100 (Feb. 11, 1955).

T. J. Balles, Coupling Between Probes Used for Flow Studies on the Metal and Plastic
Core Medels for the ART Reactor, ORNL CF-55-11-117 (Nov. 18, 1955).

J. C. Amos, Small Fluoride-NaK Heat Exchanger Test No. 1 (High Velocity Heat Ex-
changer Test), ORNL CF-56-1-187 (Jan. 2, 1956}.

J. M. Eastman, NaK Systems for Varying Reactor Load Coupling, ORNL CF-56-2-101
(Feb. 21, 1956).

C. B. Thompson, Steady-State Control Characteristics of Chemical-Nuclear Aircraft
Power Plants, ORNL-1976 {Feb. 29, 1956).

J. J. Milich, ART Fuel Dump Valve Test, ORNL CF-56-4-42 {April 6, 1956},

A. L. Southern, Closed-Loop Level Indicator for Corresive Liquids Operating at High
Temperatures, ORNL-2093 (May 17, 1956).

PHYSICS

C. B. Mills, Heating in the B ,C Curtain Due to Neutron Absorption and the B1%(n,a)Li’
Reaction, ORNL Y-F10-64 (zug 16, 1951).

C. B. Mills, A Simple Criticality Relation for Be-Moderated Intermediate Reactors,
ORNL Y-F10-93 (March 10, 1952).

W. K. Ergen, Physics Considerations of Circulating-Fuel Reactors, ORNL Y-F10-98
(April 16, 1952).

F. G. Prohammer, Note on the Linear Kinetics of the ANP Circulating-Fuel Reactor,
ORNL Y-F10-99 (April 22, 1952).

W. K. Ergen, The Kinetics of the Circulating-Fuel Nuclear Reactor, ORNLL CF-53-3-231
{March 30, 1953).

W. K. Ergen, The Inbour Formula for a Circulating-Fuel Nuclear Reactor with Slug Flow,
ORNL CF-53-12-108 (Dec. 22, 1953}.

W. K. Ergen, The Bebhavior of Certain Functions Related to the Inbour Formula of
Circulating-Fuel Reactors, ORNL CF-54-1-1 (Jan. 15, 1954).

J. L. Meem, The Xenon Problem in the ART, ORNL CF-54-5-1 (May 3, 1954).

- W. K. Ergen, The Kinetics of the Circulating-Fuel Reactor, ORNL CF-54-4~6 {May 5,
S 1954). |

177
178

J. E. Fautkner, Calculation of Fission Neutron Age in NaZrF_, ORNL CF-54-8-97
(Aug. 31, 1954).

4. E. Faulkner, Age Measurements in LiFF, ORNL CF-54-8-98 (Aug. 31, 1954).

P. H. Pitkanen, On Gamma Ray Heating in the Reflector-Moderated Reactor, ORNL
CF-54-2-111 (Sept. 14, 1954).

J. W. Noaks, Preliminary FEvaluation of Possible Poisons for Use in the ART Control
Rod, ORNL CF-55-2-16 {Feb. 2, 1955).

L. T. Anderson, Gamma and Neutron Heating of the ART Fuel Pump Assembly, ORNL
CF-55-3-161 (March 28, 1955).

L. T. Anderson, Calculation of the Beryllium Contribution to the ART Temperature
Coefficient of Reactivity, ORNL CF-55-5-76 (May 11, 1955).

W. K. Ergen, Probable Effect of Replacing Inconel by Columbium in ART Core Shells,
ORNL CF-55-5-100 (May 13, 1955).

A. M. Perry, The Boron Laver of the ART, ORNL CF-55-8-38 (Aug. 4, 1955).

W. K. Ergen, Rate of Molecular Dispersion of Heat in the ART and of Dye in an Aqueous
Solution, ORNL CF-55-9-40 (Sept. 9, 1855).

H. W. Bertini, An Estimate of the Self-Abscrption of the Decay Gammas in the ART
Fuel Dump Tank, ORNL CF-55-12-47 (Dec. §, 1955).

H. W. Bertini, Activity of the Na Cooclant in the ART, ORNL CF-55-12-78 (Dec. 16,
1955).

A. M. Perry, Fission Power Distribution in the ART, ORNL CF-56-1-172 (Jan. 25, 1956).

W, K. Ergen, Uranium Investment in a Circulating Fluoride Power-Producing Reactor,
ORNL CF-56-4-29 (April 4, 1956).

A. M. Perry, ART Pile Period During Fill Operation, ORNL CF-56-4-34 (April 9, 1956).

C. N. Copenhaver, Basic Inelastic Neutron Data for Age Calculations for Various Fuel
and Moderator Compositions, ORNL CF-56-6-40 (June 6, 1956).

A. M. Perry, Effect of Gaps in the Boron Layer, CRNL CF-56-6-65 {June 13, 1956).

A. M. Perry, Burnup of Boron in ART and in Sample lrradiations, ORNL CF-56-6-153
(June 29, 1956).

H. W. Bertini et al., Basic Gamma-Ray Data for ART Heat Deposition Calculations,
ORNL-2113 (July 5, 1956).

51’

CRITICAL EXPERIMENTS

D. Scott and B. L. Greensireet, Reflector Moderated Critical Assembly, Experimental
Program, ORNL CF-54-4-53 (April §, 1954),

D. Scott, The First Assembly of the Small Two Region Reflector-Moderated Reactor,
ORNL CF-54-7-15% (July 26, 1954},

D. Scott, The Second Assembly of the Small Two Region Reflector Moderated Reactor,
ORNL CF-54-8-180 (Aug. 25, 1954).

D. Scott and R. M. Spencer, The Second Assembly of the Small Two Region Reflector
Moderated Reactor {Part 11}, ORNL CF-54-9-185 (Sept. 27, 1954).

B. L. Greenstreet, Reflector Moderated Critical Assembly Experimental Program —
Part H, ORNL CF-54-10-119 (Oct. 19, 1954).

R. M. Spencer, The First Assembly of the Three Region Reflector Moderated Reactor,
ORNL CF-54-11-33 (Nov. 5, 1954}.

R. M. Spencer, Preliminary Critical Assemblies of the Reflector-Moderated Reactor,
ORNL-1770 (Nov. 22, 1954).

B. L. Greenstreet and R. M. Spencer, The Second Assembly of the Three Region Re-
flector Moderated Reactor, ORNL CF-54-11-150 (Nov. 24, 1954).

R. M. Spencer, Three Region Reflector Moderated Critical Assembly (Experimental
Results), ORNL CF-54-12-189 (Dec. 28, 1954).

 
 

R. M. Spencer, Three Region Reflector Moderated Critical Assembly with 146-1'720 Inconel
Core Shells, ORNL CF-55-1-123 (Jan. 21, 1955),

J. W. Noaks, Preliminary Evaluation of Possible Poisons for Use in the ART Control
Rod, ORNL CF-55-2-16 (Feb. 2, 1955).

R. M. Spencer, Three Region Reflector Moderated Critical Assembly with I,é-z'n.. Inconel
Core Shells, ORNL CF-55-2-93 (Feb. 14, 1955).

J. W. Noaks, Fvaluation of Reactivity Characteristics of Control Rods and Materials
Potentially Suitable for Use in the ART. Part I, ORNL CF-55-4-84 {April 13, 1955).
J. W. Noaks, Fvaluation of ART Control Rod Materials, Part Ill. The Effects of Neutron

Irradiation on Some Rare Earth Samples, ORNL CF-55-4-137 (April 25, 1955),

J. W. Noaks, Fvaluation of ART Control Rod Materials, Part IV. The Variation of
Reactivity with Control Rod Diameter, ORNL CF-55-4-178 (April 29, 1955).

J. W. Noaks, Evaluation of ART Control Rod Materials, Part V. Addendum to Parts
I-1v, ORNL CF-55-5-147 (May 20, 1955).

E. V. Sandin, Reflector Moderated Critical Assembly Experimental Program — Part I,
ORNL CF-55-5-181 {May 27, 1955},

D. Scott, J. J. Lynn, and E. V. Sandin, Reflector Mocderated Critical Assembly with
End Ducts — Experimental Program (Experimental Results), ORNL CF-55-6-94 {June
16, 1955).

S. Snyder et al., Three Region Reflector Moderated Critical Assembly with End Ducts
and Y -in. Inconel Core Shells. CA-21-1 Neutron Flux Measuremeni, CRNL CF-
55-10-142 (Oct. 28, 1955).

D. Scott et al., Three Region Reflector Moderated Critical Assembly with End Ducts.
Experimental Results with CA-22, Enlarged End Duct Modification, ORNL CF-56-1-96
(Jan. 30, 1956).

D. Scott et al., Three Region Reflector Moderated Critical Assembly with End Ducts —
Experimental Results with CA-23, Enlarged Island Modification, ORNL CF-56-1-97
(Jan. 30, 1956).

CORROSION

L. S. Richardson, D. C. Vreeiand, and W. D. Manly, Corrosion by Molten Fluorides,
ORNL-1491 (March 17, 1953).

G. M. Adamson, Examiration of Bi-Fluid Loop No. I, ORNL CF-53-7-199 (July 29,
1953).

H. Inolye, Scaling of Columbium in Air, ORNL-1565 (Sept. 1, 1953).

G. M. Adamson and R. S, Crouse, Examination of LF Pump Loop, ORNL CF-53-10-117
(Oct. 6, 1953).

G. M. Adamson and R, S. Crouse, Metallographic Examination of Second Heat Exchanger
from Bi-Fluid Pump Loop, ORNL CF-53-10-228 (Oct. 27, 1953).

G. M. Adamson, Metallographic Examination of Ferced Circulation Loop No. 2, ORNL
CF-54-3-15 (March 3, 1954).

W, D. Manly, High Flow Velocity and High Temperature Gradient Loops, ORNL CF-
54-3-193 (March 18, 1954).

W. D. Manly, Interim Report on Static Liquid Metal Corrosion, ORNL-1674 (May 11,
1954).

G. M.)Adamson and E. Long, Examination of Sodium, Beryllium, Inconel Pump Loops
Numbers 1 and 2, ORNL CF-54-9-98 (Sept. 13, 1954),

W. K. Stair, The Design of a Small Forced Circulation Corrosion Locp, ORNL CF-
54-10-97 (Oct. 11, 1954).

G. M. Adamson and R. 3. Crouse, Examination of First Three Large Fluoride Pump
Loops, ORNL CF-55-3-157 {March 24, 1955),

G. M. Adamson, R. S. Crouse, and P. G. Smith, Examinatiorn of Inconel-Fluoride 30-D
Pump Loop Number 4695-1, ORNL CF-55-3-179 {March 28, 1955},

179
180

G. M. Adamson and R. S. Crouse, Examination of Sodium-Inconel Pump l.oop 4689-4,
ORNL CF-55-4-167 (Aprii 21, 1955),

G. M. Adamson and R. S. Crouse, Examination of Fluoride Pump Loops 4930-A and
4935-1, ORNL CF-55-4-181 (April 26, 1955).

G. M. Adamson and R. S. Crouse, Examination of Fluoride Pump Loops Numbers 4695-2
and 4695-3, ORNL CF-55-5-97 (May 11, 1955}.

R. S. Crouse, E. Long, and A. Toboada, Examination of Sodium Inconel Beryllium
Loops 4667-3 and 4907-7, ORNL CF-55-5-123 (May 12, 1955},

G. M. Adamseon and R. S. Crouse, Examination of Inconel-316 Stainless Steel—Sodium
Pump Loops 4689-5 and 4689-6, ORNL CF-55-6-24 (June 2, 1655).

G. M. Adamson and R. S. Crouse, Effect of Heating Method on Fluoride Corrosion.
Examination of Fluoride Pump Loops 4935-2, 4950-1, 4950-2, ORNL CF-55-6-99
(June 14, 1955).

R. J. Gray, Evidences of Mass Transferred Material in Inconel and 316 Stainless Steel
Sodium-to-Air Radiator, CRNL CF-55-6-89 (June 15, 1955).

G. M. Adamson and R. S. Crouse, Examination of Short-Operating Time Loops Numbers
4695-4 and 4695-5, ORNL CF-55-7-66 (July 8, 1955),

G. M. Adamson and R. §. Crouse, Examination of Loop 4950-3-Inconel with a UF and
UF ; Mixture in Zirconium Fluoride, ORNL CF-55-8-70 (Aug. 9, 1955).

G. M. Adamson and R. S. Crouse, Examination of Inconel-Fluoride Pump Loops 4935-3
and 4935-4, ORNL CF-55-8-94 (Aug. 12, 1955).

G. M. Adamson and R. S. Crouse, Examination of Final Loops in Time Series Numbers
4695-4D(2) and 4695-5C(2;, ORNL CF-55-8-151 (Aug. 19, 1955).

R. 5. Crouse, J. H. DeVan, and E. A. Kovacevich, Examination of Fluoride Pump Loops
4935-5 and 4935-7, ORNL CF-55-10-31 (Oct. 5, 1955).

R. S. Crouse, J. H. DeVar, and E. A, Kovacevich, Examination of Miscellaneous Pump
Loops 4950-4 and 4917, ORNL CF-55-10-32 (Oct. 5, 1955),

R. J. Gray and P. Patriarca, Metallographic Examination of ORNL Radiator No. 1 and
York Radiator No. 1 Failures, ORNL CF-55-10-129 (Oct. 31, 1955).

H. lnouye, M. D’Amore, and J. H. Coobs, Boron Containing Materials — Inconel Com-
patibility, ORNL CF-55-12-117 (Dec. 22, 1955).

J. V. Cathcart and W. D. Manly, The Mass-Transfer Properties of Various Metals and
Alloys in Liguid Lead, ORNL-2008 {Jan. 10, 1956).

J. H. DeVan, Examination of SHF-C Resistance Heater, ORNL CF-55-5-145 (May 31,
1956).

J. V. Catheart, L. L. Hall, end G. P. Smith, Oxidation Characteristics of the Alkali
Metals; 1. Oxidation Rate of Sodium Between ~79 and 48°C, ORNL-2054 (May 22,
1956).

R. J. Gray, Metallographic Examination of High Velocity Heat Fxchanger (SHE No. 1),
ORNL CF-56-5-148 (May 24, 1956).

T. Hikido, The Role of Chromium in the Mass Transfer Attack of Alloys by Fluorides,
ORNL CF-56-6-58 (June 10, 1956).

J. H. DeVan and R. 5. Crouse, Examination of Hastelloy B Pump Loops 7425-12 and
-13, ORNL CF-56-6-79 {June 11, 1956).

J. H. DeVan and R. S. Crouse, Examination of Pump Loops 7425-14 and 7425-15 Which
Circulated Fuel 70, ORNL CF-86-6-161 (June 29, 1956).

J. H. DeVen, Effect of Temperature on Depth of Attack in Inconel-Fuel 30 Pump Loops,
ORNL CF-56-7-9 (July 5, 1956).

W. D. Merly, Fundamentals of Liguid-Metal Corrosion, ORNL-2055 (July 12, 1956},

J. E. Van Cleve, Metallographic Examination of I.LH.E, No. 3, ORNL CF-554-7-82 (July
12, 1956).

 
J. H. DeVan and R. S. Crouse, Examination of ORNL [ and 2 Intermediate Heat Ex-
changers, Type IHE-3, ORNL CF-56-7-135 (July 20, 1956).

J. E. Van Cleve, Metallographic Examination of §.H.E. No. 2 (ORNL No. }) Heat Ex-
changer, ORNL CF-56-7-130 (July 27, 1956)}.

CHEMISTRY

T. N. McVay, Petrographic Examination of Fluoride Fuels, ORNL. CF-52-6-127 (June 20,
1952).

C. J. Barten, Fused Salt Compositions, ORNL CF-53-1-129 (Jan. 15, 1953).

C. J. Barton, Fused Salt Compositions, ORNLL CF-53-10-78 {Oct. 9, 1953),

E. Orban, Data on the BeF ,-NaF System, ORNL CF-54-4-47 (April 27, 1954),

R. F. Newton, Effects of Fission Froducts on Performance of a Reactor Using Fluorides
as Solvent for Fuel, ORNL CF-54-5-40 (May 7, 1954).

F. F. Blankenship, Analysis of Salt Mixtures, ORNL CF-54-5-189 (May 25, 1954).

E. Orban, Measurement of the Stability of Lithium Hydride, ORNL CF-54-5-47 (May 286,
1954).

C. J. Barton, Fused Salt Compositions, ORNL CF-54-6-6 (June 2, 1954),

J. C. White, Sclubility of Composition 30 in Water, ORNL CF-55-4-18 {April 4, 1955),

J. C. White, A. S. Meyer, Jr., and D. L. Manning, Differential Spectrophotometric De-
termination of Beryllium, ORNL-1909 (June 24, 1955),

J. C. White, Determination of Iron(lll) in Mixiures of Alkali Metal Fluoride Salt, ORNL
CF-55-7-103 (July 22, 1955}.

C. J. Barton, Fused Salt Compositions, ORNL CF-55-9-78 (Sept. 16, 1955),

J. C. White, Determination of Traces of Iron(iIl) in NaFaLiF-KF-UF4 with Thiocyanate,
ORNL CF-55-9-96 (Sept. 20, 1955).

D. L. Manning, A. S. Meyer, Jr., and J. C. White, The Compleximetric Titration of
Zirconium Based on the Use of Ferric Iron as ithe Titrant and Disodium-i, 2-Diby-
droxybenze-3, 5-Disulfonate as the Indicator, ORNL-1950 {Sept. 28, 1955).

W. J. Ross, A. S. Meyer, Jr., and J. C, White, Determination of Trivalent Uranium with
Methylene Blue, ORNL-1984 (Nov. 7, 1955).

J. C. White, Chemical Examination of Cold Trap from Intermediate Heat Exchanger Test
Stund No. 1, ORNL CF-55-11-102 (Nev. 17, 1955).

J. C. White, Determination of Small Amounts of Tantalum in NaF-LiF-KF and in NaF-
LiF-KF-UF ., ORNL CF-56-1-49 {Jan. 10, 1956).

J. R. Smolen, Physical Properties of Boral Used in Lid Tank Mockups, ORNL CF-
56-1-94 (Jan. 23, 1956).

J. C. White ez al., Determination of Trivalent Uranium in Fluoride Sait Mixtures by the
Modified Hydrogen Evolution Method, ORNI.-2043 (Feb. 10, 1956).

J. C. White and A. S. Meyer, Determination of Traces of Aluminum in NaF-ZrF -UF ,,
ORNL CF-56-3-10 (March 1, 1956).

J. C. White, A. $. Meyer, and B. L. McDowell, Determination of Dissolved Oxygen in
Lubricating Fluids, ORNL-2059 (March ¢, 1956).

J. C. White, Procedure for the Determination of Oxygen in Sodium and NaK by the
Distillation Metbod, ORNL CF-56-4-31 (April 5, 1956).

P. A. Agron, B. S. Borie, Jr., and R. M. Steele, 83°1\‘T‘I2UF6’ ORNL CF-56-4-199 (April
30, 1956),

R. E. Thoma, X-ray Diffraction Results, ORNL CF-56-6-25 {June 4, 1956},

J. C. White, Determination of Microgram Amounts of Titanium in the Presence of NaF-
ZrF ;-UF ,, ORNL CF-56-6-111 (June 20, 1956).

W. J. Ross, A, S. Meyer, Jr., and J. C. White, Determination of Boron in Fluoride Salts,
ORNL-2135 (Aug. 7, 1956).

181
182

METALLURGY

A. DeS. Brasunas, A Simplified Apparatus for Making Thermal Gradient Dynamic Cor-
rosion Tests (Seesaw Tests), ORNL CF-52-3-123 (March 13, 1952).

W. D. Manly, Status of Columbium, Beryllium, and Lithium, ORNL CF-54-4-162 (April 6,
1954).

E. E. Hoftman and P. Patriarca, Heat Exchanger Fabrication, ORNL CF-54-5-88 (May
17, 1954).

G. P. Smith, M. E. Steidlitz, and L. L. Hall, Flammability of Sodium Alloys at High
Temperatures, ORNL-1769 (June 15, 1955).

H. lnouye, Neutron Shieid for ART, ORNL CF-55-7-94 (July 18, 1955).

R. J. Gray and P. Patriarca, Metallographic Examination of ORNL HCF Radiator No. |
Failures, ORNL CF-55-10-129 (Oct. 31, 1955).

H. inouye, Use of Boror for Neutron Skielding, ORNL CF-55-12-3 {Dec. 1, 1955).

R. S. Crouse, Metallographic Examination of Thermal Convection Loops and Capsule
Tests Performed by WADC, ORNL CF-56-2-26 (Feb. 6, 1956).

J. H. DeVan, Preliminary Investigation of Inconel Castings, ORNL CF-56-2-56 (Feb. 13,
1956).

R. J. Gray and P. Petriarca, Metallographic Examination of PWA HCF Radiator No. 2,
ORNL CF-56-3-47 (March 12, 1956},

R. Carlander and E. E. Hoffman, Transfer of Carbon Between Dissimilar Metals in
Contact with Molten Sodium, ORNL CF-56-4-73 (April 2, 1956).

H. Inouye, M. D’Amore, and T. K. Roche, Nickel Base Alloys for High Temperature
Service, ORNL CF-56-4-121 (April 16, 1956).

M. R. D’Amore and H. lnouye, The Extrusion of Composite Tubes, ORNL CF-54-4-123
(April 18, 1956).

R. J. Gray, Metallographic Examination of High Velocity Heat Exchanger (SHE No. 1),
ORNL CF-56-5-148 (May 24, 1956).

P. Patriarca et al., Fabrication of Heai Exchangers and Radiators for High Temperature
Reactor Applications, ORNL-1955 (June 14, 1955),

T. Hikide, Screening Evaluation of Experimental Nickel-Molybdenum Alloys, ORNL
CF-56-7-35 (July 10, 1956).

HEAT TRANSFER

H. C. Claiborne, A Critical Review of the Literature on Pressure Drop in Noncircular
Ducts and Annuli, ORNL-1248 (May 6, 1952),

H. F. Poppendiek and L. D. Palmer, Forced-Convection Heat Transfer in Thermal
Entrance Regions — Part II, ORNL-914 (May 26, 1952).

W. S. Farmer, Cooling Hole Distribution for Reactor Reflectors, ORNL CF-52-9-201
(Sept. 3, 1952).

W. S. Farmer, Heat Generation in a Slab for a Non-Uniform Source of Radiation, ORNL
CF-52-9-202 (Sept. 4, 1952},

H. F. Poppendiek, Estimates of Heat and Momentum Transfer Characteristics of the
Two Fluoride Coolants: (LiF-48 M %, BeFZ-_SZ M %) and (NaF-10 M %, KF-46 M %,
ZrF ;-44 M %), ORNL CF-52-11-205 (Nov. 29, 1952).

H. F. Poppendiek and L. D. Palmer, Forced-Convection Heat Transfer in Pipes with
Volume Heat Sources, ORNL-1395 (Dec. 2, 1952).

W. B. Harrison, J. O. Bradfute, and R. V. Bailey, Generalized Velocity Distribution for
Turbulent Flow in Annuli, ORNL CF-52-12-124 (Dec. 19, 1952).

W. B. Harrison, Forced Convection Heat Transfer in Thermal Entrance Regions, Part 11,
ORNL-915 (Aug. 6, 1953).

H. W. Hoffman, Preliminary Results of Flinak Heat Transfer, ORNL CF-53-8-106 {Aug.
18, 1953).

 
H. F. Poppendiek and G. M. Winn, Some Preliminary Forced-Convection Heat Transfer
Experiments in Pipes with Volume Heat Sources Within the Fluids, ORNL CF-54-2-1
{Feb. 1, 1954).

J. O. Bradfute, The Measurement of Fluid Velocity by a Photographic Technique, ORNL
CF-54-2-37 (Feb. 4, 1954),

D. C. Hamilton, F. E. Lynch, and L. D. Palmer, The Nature of the Flow of Ordinary
Fluids in a Thermal-Convection Harp, ORNL-1624 (Feb. 23, 1954),

L. D. Palmer and G. M. Winn, A Feasibility Study of Flow Visualization Using a Phos-
phorescent Particle Method, ORNL CF-54-4-205 {(April 30, 1954).

H. F. Poppendiek and L. D. Palmer, Forced Convection Heat Transfer Between Parallel
Plates and in Annuli with Volume Heat Sources Within the Fluids, ORNL-1701 (May 11,
1954).

M. W. Rosenthal, H. F. Poppendiek, and M. R. Burnett, A Method for Evaluating the
Heat Transfer Effectiveness of Reactor Coolants, ORNL CF-54-11-63 {Nov. 4, 1954).
H. F. Poppendiek, A Laminar Forced-Convection Solution for Pipes Ducting Liguids
Having Volume Heat Sources and lLarge Radial Differences in Viscosity, ORNL

CF-54-11-37 (Nov. 5, 1954).

D. C. Hamilton et al., Free Convection in Fluids Having a Volume Heat Source, ORNL-
1769 (Nov. 15, 1954),

J. O. Bradfute, Qualitative Velocity Information Regarding the ART Core: Status
Report No. 4, ORNL CF-54-12-110 (Dec. 14, 1954).

H. W. Hoffman and J. Lones, Fused Salt Heat Transfer — Part ' Forced Convection
Heat Transfer in Circular Tubes Containing NaF-KF-LiF Eutectic, ORNL-1777 (Feb.
1, 1955).

H. W. Hoffman, i.. D. Palmer, and N, D. Greene, Electrical Heating and Flow in Tube
Bends, ORNL CF-55-2-148 (Feb. 22, 1955).

G. L. Muller and J. O. Bradfute, Qualitative Velocity Profiles with Rotation in 18-Inch
ART Core, ORNL CF-55-3-15 (March 1, 1955).

H. F. Poppendiek, Status Report on Forced Convection Experimental Work in Converging
and Diverging Channels with Volume Heat Sources in the Fluids, ORNL CF.55-3-174
{(March 24, 1955).

F. E. Lynch and J. O. Bradfute, Qualitative Velocity Profiles in 18-Inch ART Core
with Increased Turbulence at Its Iniet, ORNL CF-55-5-132 (May 10, 1955).

J. O. Bradfute, F. E. Lynch, and G. L. Muller, Fluid Velocity Measured in the ]8-Inch
ART Core by a Particle-Photographic Technigue, ORNL CF-55-6-137 (June 21, 1955),
F. E. Lynch, Qualitative Velocity Information Regarding the ]8+Inch ART Core with

Vane Set No. 2 in Entrance, ORNL CF-55-6-173 (June 27, 1955},

G. L. Muller, Qualitative Velocity Information Regarding the 18-Inch ART Core with
Turbulator Vane Set No. I at Entrance, ORNL CF-55-4-174 (June 27, 1955).

L. D. Palmer, G. M. Winn, and H. F. Poppendiek, Investigation of Fluid Flow in Helical
Bends of the ANP In-Pile Loop, ORNL CF-55-6-183 (June 27, 1955).

G. L. Muller, Qualitative Estimates of the Velocity Profiles in the 21-Inch ART Core,
ORNL CF-55-7-92 (July 20, 1955).

D. C. Hamilton and F. E. Lynch, Free Convection Theory and Experiment in Fluids
Having a Volume Heat Source, ORNL-1888 (Aug. 3, 1955).

H. F. Poppendiek and L. D. Palmer, Application of Temperature Solutions for Forced
Convection Systems with Volume Heat Sources to General Convection Problems,

ORNL-1933 (Sept. 29, 1955).
G. L. Muller and F. E. Lynch, Qualitative Velocity Information Regarding the Quarter

Scale Model of the 18-Inch ART Core and the 5.22 Scale Model of the 2i-Inch ART
Core with the P and W Swirl Nozzles at the Inlet, QORNL CF-55-10-48 (Oct. 3, 1955),

183
184

G. L. Muller and F. E. Lynch, Qualitative Velocity Information Regarding Two Constant
Gap Core Models, ORNL CF-55-10-49 (Oct. 3, 1955).

F. E. Lynch and G. L. Muller, Qualitative Velocity Profiles with a Rotational Com-
ponent at the Inlet of the 2i-Inch Core Model, G. F. Wislicenus Design, ORNL CF-
55-10-50 (Cct. 3, 1955),

4. L. Wantland, Therma! Characteristics of the ART Fuel-to-NaK Heat Exchanger,
ORNL. CF-55-12-120 (Dec. 22, 1955),

J. L. Wantlond, A Method of Correlating Experimental Fluid Friction Data for Tube
Bundles of Different Size and Tube Bundle to Shell Wall Spacing, ORNL. CF-56-4-162
(April 5, 1956).

H. W. Hoffman, Thermal Structure for the Region Beyond the ART Reflector — Supple-
ment 1, CRNL CF-56-4-129 (April 17, 1956).

H. W. Hoffman, Fused Salt Heat Transfer, ORNL CF-56-7-85 (July 20, 1956).

PHYSICAL PROPERTIES

M. Tobias, S. |. Kaplan, and S. J. Claiberne, Densities of Certain Salt Mixtures at
Room Temperature, ORNL CF-52-3-230 (March 26, 1952).

J. M. Cisar, Densities of Fuel Mixtures Nos. 25 and 25a, ORNL CF-52-5-209 (May 28,
1952).

R. F. Redmond and T. N. Jones, Viscosity of Fulinak, ORNL CF-52-4-148 (June 19,
1952).

R. F.)Redmond and T. N. Jones, Viscosity of Fuel Sait Mixtures No. 27 and No. 30,
ORNL. CF-52-7-138 (July 23, 1952).

L. Cooper and S. J. Claiborne, Measurement of the Thermal Conductivity of Flinak,
ORNL CF-52-8-163 (Aug. 26, 1952).

W. B. Harrison, Transient Methods for Determining Thermal Conductivities of Liguids,
ORNL CF-52-11-113 (Nov. 1, 1952),

S. J. Claiborne, Measurement of the Thermal Conductivity of Fluoride Mixture No. 35,
ORNL CF-52-11-72 (Nov. 11, 1952).

W. D. Powers and G. C. Blalock, Heat Capacity of Fused Salt Mixture No. 21, ORNL
CF-52-11-103 (Nov. 15, 1952).

R. F. Redmond and T. N. Jones, Density Measurements of Fuel Salts Nes. 31 and 33,
ORNL CF-52-11-105 {Nov. 15, 1952).

R. F. Redmond and T. N. Jones, Viscesity Measurements of Fuel Salts Nos. 31 and 33,
ORNL CF-52-11-106 {(Nov. 15, 1952).

S. J. Claiborne, Measurement of the Thermal Conductivity of Fluoride Mixtures No. 30
and No. 14, ORNL CF-53-1-233 (Jon. 8, 1953).

R. F. Redmond and S. |. Kaplan, Remarks on the Falling-Ball Viscometer, ORNL CF-
53-1-248 (Jan. 14, 1953).

W. D. Powers and G. C. Blalock, Heat Capacity of Fused Salt Mixture No. 30, ORNL
CF-53-2-56 (Feb. 6, 1953},

Pbysical Properties Charts for Some Reactor Fuels, Coolants, and Miscellaneous
Materials, 3rd ed., ORNL CF-53-3-261 (March 20, 1953),

W. D. Powers and G. C. Blalock, Heat Capacity of Fuel Composition No. 14, ORNL
CF-53-5-113 {May 18, 1953).

W. J. Sturm, R. J. Jones, and M. J. Feldman, The Stability of Several Fused Salt
Systems Under Proton Bombardment, ORNL-1530 (June 19, 1953).

S. i. Cohen and T. N. Jones, Preliminary Measurements of the Density and Viscosity
of Fluoride Mixture No. 40, ORNL CF-53-7-125 (July 23, 1953).

S. I. Cohen and T. N. Jones, Measurements of the Solid Densities of Flucride Mixture
No. 30, Ber, and NaBeF ,, ORNL CF-53-7-126 (duly 23, 1953).

W. D. Powers and G. C. Blalock, Heat Capacity of Fuel Composition No. 12, ORNL
CF-53-7-200 (July 31, 1953).
S. I. Cohen and T. N. Jones, Preliminary Measurements of the Density and Viscosity
of Fluoride Mixture No. 44, ORNL CF-53-8-217 (Aug. 31, 1953).

S. |. Cohen and T. N. Jones, Preliminary Measurements of the Density and Viscosity
of Composition 43 (Na,UF ), ORNL CF-53-10-86 (Oct. 14, 1953).

W. D. Powers and G. C. Blalock, Heat Capacity of Fuel Composition No. 33, ORNL
CF-53-11-128 (Nov. 23, 1953).

S. I. Cohen and T. N. Jones, Preliminary Measurements of the Density and Viscosity
of NaFoZrF4=UF (62.5-12.5-25.0 mole %), ORNL CF-53-12-179 (Dec. 22, 1953).

W. D. Powers anfi G. C. Blalock, Heat Capacity of Fuel Composition No. 31, ORNL
CF-54-2-114 (Feb. 17, 1954).

S. I. Cohen and T. N. Jones, A Summary of Density Measurements on Molten Fluoride
Mixtures and a Correlation Useful for Predicting Densities of Fluoride Mixtures of
Known Compositions, ORNL-1702 (May 14, 1954).

W. D. Powers and G. C. Blalock, Heat Capacities of Composition No. 12, No. 44, and
of K;CrF ., ORNL CF-54-5-160 (May 20, 1954).

S. I. Cohen et al., Pbysical Property Charts for Some Reactor Fuels, Coolants, and
Miscellanecus Materials (4th Edition), ORNL CF-54-5-188 (June 21, 1954).

S. I, Cohen and T. N. Jones, Measurement of the Density of Ligquid Rubidium, ORNL
CF-54-8-10 (Aug. 4, 1954).

W. D. Powers and G. C, Blalock, Heat Capacities of Compositions Nos. 39 and 101,
ORNL CF-54-8-135 (Aug. 17, 1954).

W. D. Powers and S. J. Claiborne, Measurement of the Thermal Conductivity of Molten
Fluoride Mixture No. 104, ORNL CF-54-7-145 {(Aug. 24, 1954).

W. D. Powers and S. J. Claiborne, Measurement of the Thermal Conductivity of Molten
Fluoride Mixture No. 44, ORNL CF-54-10-139 (Ocr. 26, 1954).

W. D. Powers and G. C. Blalock, Heat Capacity of Composition No. 40, ORNL CF-
54-10-140 (Oct. 26, 1954).

S. I. Cohen and T. N. Jones, Preliminary Measurements of the Viscosity of Composition
20, ORNL CF-55-2-20 (Feb. 2, 1955).

S. |. Cohen and T. N. Jones, Measurement of the Viscosities of Composition 35 and
Composition 74, ORNL CF-55-2-89 (Feb. 15, 1955).

W. D. Powers and G. C. Blalock, Heat Capacity of Lithium Hydride, ORNL CF-55-3-47
(March 7, 1955),

S. I. Cohen and T. N. Jones, Measurement of the Viscosity of Composition 72, ORNL
CF-55-3-61 (March 8, 1955).

S. . Cohen and T. N. Jones, Measurement of the Viscosity of Composition 30, ORNL
CF-55-3-62 (March 9, 1955).

S. I. Cohen and T. N. Jones, Measurement of the Viscosity of Composition 43, ORNL
CF-55-3-137 (March 14, 1955),

S. I. Cohen and T. N. Jones, Measurement of the Viscosity of Composition 2, ORNL
CF-55-4-32 {April 1, 1955).

S. I. Cohen and T. N. Jones, Measurement of the Viscosity of Compositions 81 and 82,
ORNL CF-55-5-58 (May 16, 1955).

S. |. Cohen and T. N. Jones, Measurement of the Viscosity of Composition 78, ORNL
CF-55-5-59 (May 16, 1955).

W. D. Powers and G. C. Blalock, Heat Capacities of Compositions 30, 31, and 40,
ORNL CF-55-5-87 (May 18, 1955).

W. D. Powers and G. C. Blaleck, Heat Capacities of Compositions 70 and 103, ORNL
CF-55-5-88 {(May 18, 1955). :

S. |. Cohen and T. N. Jones, Measurement of the Viscosity of Composition 86, ORNL
CF-55-7-33 (July 7, 1955},

H. W. Hoffman, Physical Properties and Heat Transfer Characteristics of an Alkali
Nitrate-Nitrite Salt Mixture, ORNL CF-55-7-52 {July 21, 1955).

 

185
185

W. D. Powers and G. C. Blolock, Heat Capacity of Composition 102, ORNL CF-55-5-8
(Aug. 1, 1955).

S. I. Cohen and T. N. Jones, Measurement of the Viscosity of Composition 88, ORNL
CF-55-8-21 (Aug. 1, 1955},

S. |. Cohen and T. N. Jones, Measurement of the Viscosities of KBeF, and NaBng
and Some Observations on (LiF-BeF,, 50-50 mole %), ORNL CF-55-8-22 (Aug. 1,
1955).

S. i, Cohen and T. N. Jones, Measurement of the Viscosity of Composition 70, ORNL
CF-55-9-31 (Sept. 6, 1955).

S. |. Cohen and T. N. Jones, Measurement of tke Viscosity of Compositions 87, 95,
and 104, ORNL CF-55-11-27 (Nov. 4, 1955),

S. |. Cohen and T. N. Jones, Measurement of the Viscosity of Compositions 90 and 96
and (KE-BeF,; 79-21 mole %}, ORNL CF-55-11-28 (Nov. 8, 1955).

W. D. Powers and G. C. Blalock, Heat Capacities of Compositions No. 98 and No. 104,
ORNL CF-55-11-68 {Nov. 15, 1955).

S. I. Cohen and T. N. Jones, Measurement of the Viscosity of Compositions 97 and 98,
ORNL CF-55-12-127 {Dec. 23, 1955).

S. |. Cohen and T. N. Jones, Measurement of the Viscosity of Composition 31, ORNL
CF-55-12-128 (Dec. 23, 1955).

W. D. Powers and G. C. Blalock, Enthalpies and Heat Capacities of Sclid and Molten
Fluoride Mixtures, ORNL-1956 (Jan. 11, 1956).

S. |. Cohen and T. N. Jones, The Effect of Chemical Purity on the Viscosity of a
Molten Fluoride Mixture, ORNL CF-56-4-148 (April 17, 1956).

S. I. Cohen and T. N. Jones, Measurement of the Viscosity of Compositions 12, 14, and
107, ORNL CF-56-5-33 (May G, 1956).

W. D. Powers and G. C. Blalock, Heat Capacities of Several Scdium Fluoride, Zirconium
Fluoride Compositions, ORNL CF-56-5-66 {May 14, 1956).

W. D. Powers and G, C. Blalock, Heat Capacity of Composition No. 100, ORNL CF-
56-5-67 (May 14, 1956).

W. D. Powers end G. C. Blalock, Heat Capacity of Composition No. 81, ORNL CF-
56-5-68 (May 14, 1956).

RADIATION DAMAGE

O. Sisman, LITR Fluoride Fuel Test Loop, ORNL CF-53-9-180 (Sept. 25, 1953).

W. E. Brundage and W. W. Parkinson, The Radiation-induced Corrosion of Beryllium
Oxide in Sodium at 1500°F, ORNL CF-53-12-42 {Dec. 3, 1953).

H. J. Stumpf and B. M. Wiiner, Radiation Damage to Elastomers, Lubricants, Fabrics,
and Plastics for Use in Nuclear-Powered Aircraft, ORNL CF-54-4-221 {April 15, 1954).

M. T. Robinson, Release of Xenon from Fluoride Fuels: Proposal for an Experimenial
Program, CRNL CF-54-6-4 (June 2, 1954).

M. T. Robinson, Removal of Xenon from Fluoride Fuels: Preliminary Design of In-Pile
Equipment, ORNL CF-54-9-36 (Sept. 3, 1954).

M. T. Robinson, Volatilization of Fission Products from Fluoride Fuels, ORNL CF-
541265 (Dec. 10, 1954},

M. T. Robinsen, 5. A. Reynolds, and H. W. Wright, The Fate of Certain Fission Products
in the ARE, ORNL CF-55-2-36 (Feb. 7, 1955).

M. T. Robinson, A Theoretical Treatment of Xe!33 Poisoning in the ARE and the ART,
ORNL CF-55-5-22 (May 2, 1955). .

M. T. Robinsen, Deposition of Fission Products from Fluoride Fuels, ORNL CF-55-5-99
{May 16, 1955),

M. J. Feldman et al., Metallographic Analysis of Fuel Loop I, ORNL CF-55-4-22
(June 21, 1955).

 
 

M. T. Robinson, The Fate of Fission Productis in ANP Fluoride Fuels, ORNL CF-
55-7-12 (July 1, 1955),

M. T. Robinson and D. F. Weekes, Design Calculations for a Miniature High-Temperature
in-Pile Circulating Fuel Loop, ORNL-1808 (Sept. 19, 1955).

M. J. Feldman, E. J. Mantheos, and A. E. Richt, Metallographic Analysis of Fuel Loop
II — Control Sections from Nose Piece, ORNL CF-55-9-91 (Sept. 26, 1955).

M. T. Robinson, A Theoretical Study of Xel33 Poisoning Kinetics in Fluid-Fueled,
Gas-Sparged Nuclear Reactors, ORNL-1924 (Feb. 6, 1956).

M. T. Robinson and J. F. Krause, The Use of Li in In-Pile Corrosion Testing of ANP
Fuels, ORNL CF-56-2-25 (Feb. 7, 1956). |

C. Ellis et al., Examination of ANP In-Pile Loop 5, ORNL CF-56-7-48 (JUIY 16, 1956).

FUEL RECOVERY

C. F. Coleman, Recovery of Uranium as a Single Product from Fluorides, ORNL-1500
(March 31, 1953).

A. T. Gresky, E. D, Arnold, and R. J. Klotzbach, Potentialities of Fused Salt—Fluoride
Volatility Methods of Processing Irradiated Fuel, ORNL-1803 (March 28, 1956).

SHIELDING

Report on the ANP Shielding Board, ANP-53 (Oct. 16, 1950).

W. .. Wilson and A. L. Walker, An Aircraft Shield Analysis, ORNL CF-52-8-120 (Aug.
20, 1952).

E. P. Blizard, Materials Research for Mobile Reactor Shielding, ORNL CF-52-11-129
(Nov. 18, 1952).

J. B. Trice, Summary of Results from Recent Fireball Shielding Studies, ORNL CF-
53-6-165 (June 11, 1953).

J. E. Faulkner, Critigue of LiH as a Neutron Shield, ORNL CF-53-12-13 (Dec. 2, 1953).

A. P. Froas, Shield Designs for the Reflector-Moderated Reactor, ORNL CF-54-2-36
(Feb. 4, 1954).

F. C. Maienschein, F. Bly, and T. A. Love, Sources and Attenuation of Gamma Radiation
from a Divided Aircraft Shield, ORNL-1714 {Aug. 20, 1954).

R. M. Spencer and H. J. Stumpf, The Effect on the RMR Shield Weight of Varying the
Neutron and Gamma Dose Components Taken by the Crew and Comparison of the RMR
Shield Weight to That for an Idealized Shield, ORNL CF-54-7-1 (Aug. 26, 1954).

F. N. Watson et al., Lid Tank Shielding Tests of the Reflector-Moderated Reactor,
ORNL-1616 (Oct. 5, 1954).

E. P. Blizard, Fraction of Biological Dose Due to Thermal Neutrons in Aircraft Reactor
Shields, ORNL CF-54-11-113 (Nov. 19, 1954).

J. B. Dee, Jr., and H. C. Woodsum, Shield Weights for the CFRE, ORNL. CF-54-12-154
(Dec. 20, 1954).

E. P. Blizard, Spectrometer Measurements of Fission-Product Gamma Rays for the CFR,
ORNL CF-55-4-122 (April 21, 1955).

J. Van Hoomissen and H. E. Stern, Comparison of Hydrogenous Materials for Use in
Aircraft Shields, ORNL CF-55-6-4Q (June 9, 1955},

J. B. Dee, C. A. Goetz, and J. R. Smoclen, Sources of Radiation in a 300-Mev Circu-
lating-Fuel Reactor, ORNL CF-55-7-35 (July 12, 1955).

J. R. Smolen, Pbysical Properties of Boral Used in Lid Tank Mockups, ORNL CF-
56-1-94 (Jan. 23, 1956).

C. D. Zerby, A Monte Carlo Method of Calculating the Response of a Point Detector
at an Arbitrary Position Inside a Cylindrical Shield, ORNL-2105 (June 12, 1956).

J. R. Smolen, A Preliminary Investigation of B1® Enriched Boron Curtains, ORNL CF-
56-6-163 {June 28, 1956).

First Semiannual ANP Shielding Information Meeting, May 7 and 8, 1956, Vols I Through
111, ORNL-2115 (July 2, 1956),

187
188

ORNL-528
ORNL-629
ORNL-768
ORNL-858
ORNL-919
ANP-60
ANP-65
ORNL-1154
ORNL-1170
ORNL-1227
ORNL-1294
CRNL-1375
ORNL-143¢9
ORNL-1515
ORNL-1556
ORNL-1609
ORNL-1649
ORNL-1692
ORNL-1729
ORNL-1771
ORNL-1816
ORNL-1864
ORNL-1896
ORNL-1647
ORNL-2012
ORNL-2061
ORNL-2106
ORNL-2157

AIRCRAFT NUCLEAR PROPULSION PROJECT QUARTERLY PROGRESS REPORTS

Period Ending November 30, 1949
Period Ending February 28, 1950
Feriod Ending May 31, 1950
Period Ending August 31, 1950
Period Ending December 10, 1950
Period Ending March 10, 1951
Period Ending June 10, 1951
Period Ending September 10, 1951
Period Ending December 10, 1951
Period Ending March 10, 1952
Period Ending June 10, 1952
Period Ending September 10, 1952
Periocd Ending December 10, 1952
Period Ending March 10, 1953
Period Ending June 10, 1953
Period Ending September 10, 1953
Period Ending December 10, 1953
Period Ending March 10, 1954
Feriod Ending June 10, 1954
Period Ending September 10, 1954
Period Ending December 10, 1954
Period Ending March 10, 1955
Period Ending June 10, 1955
Period Ending September 10, 1955
Period Ending December 10, 1955
Period Ending March 10, 1956
Period Ending June 10, 1956
Period Ending September 10, 1956