Program Administration and Installation Design of the Nuclear Reactor Project at North Carolina State College
Typescript
81 pp.
July 5, 1950
NEprog070550


[page i]

CONFIDENTIAL

PROGRAM ADMINISTRATION AND INSTALLATION DESIGN
OF THE
NUCLEAR REACTOR PROJECT
AT
NORTH CAROLINA STATE COLLEGE

July 5, 1950

Clifford K. Beck
and
Arthur C. Menius (on Theoretical Calculations)

George N. Webb
Arthur W. Waltner (on Instrumentation)

P. B. Leonard
F. H. Stinson
J. D. Paulson (on Drafting)

RESTRICTED DATA
This document contains restricted data as defined in the Atomic Energy Act of 1946.

CAUTION
This document contains information affecting the National Defense of the United
States
. Its transmission or the disclosure of its contents in any manner to an
unauthorized person is prohibited and may result in severe criminal penalties
under applicable Federal laws.

CONFIDENTIAL


[page ii]

CONFIDENTIAL

PROGRAM ADMINISTRATION AND INSTALLATION DESIGN
OF THE
NUCLEAR REACTOR PROJECT
AT
NORTH CAROLINA STATE COLLEGE

DISTRIBUTION OF REPORT
Series A - 25 Copies
68 Pages: i to vi and 1-62; 13 Figures

Copy No.
U.S.A.E.C. Joseph Platt1 - 6
U.S.A.E.C. Herman Roth7 - 10
U.S.A.E.C. Declassification office11 - 16
Dean J. H. Lampe17 - 18
O.R.N.L. Richard Stevenson19 - 20
Clifford K. Beck20 - 25

CONFIDENTIAL


[page iii]

CONFIDENTIAL

PROGRAM ADMINISTRATION AND INSTALLATION DESIGN
OF THE
NUCLEAR REACTOR PROJECT
AT
NORTH CAROLINA STATE COLLEGE

CONFIDENTIAL


[page iv]

CONFIDENTIAL

TABLE OF CONTENTS (Continued)

Figure1. Topographic map of Raleigh
2. N. C. State College Campus
3. Schematic Floor Plan of Reactor Building
4. Reactor Shielding Skeleton
5. Complete Reactor Shielding Assembly
6. Fuel Container
7. Reactor Envelope and Sampling Tube
8. Horizontal Cross Section through Thermal Column
9. Exposure Ports
10. Liquid Level Indicator
11. Decay of Reactor Product Gases
12. Gas Disposal System - Schematic Diagram
13. Five Second Rod Removal Emergency

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[page v]

CONFIDENTIAL

PROGRAM ADMINISTRATION AND INSTALLATION DESIGN
OF THE
NUCLEAR REACTOR PROJECT
AT
NORTH CAROLINA STATE COLLEGE

ABSTRACT

North Carolina State College of The University of North Carolina has pro-
posed
that a "water-boiler" nuclear reactor be built and operated on the college
campus as an unclassified tool of instruction and research. A brief statement of
the reactor project objectives and administrative procedure is made in this report.

The main body of the report concerns the general design and arrangement of
the reactor. Most of the design has been based on ideas and extension of ideas
and emperical data from Los Alamos and Oak Ridge reactors. The Los Alamos "water-
boiler
" has operated entirely satisfactorily for several years, and the Oak Ridge
"Homogeneous Reactor", though not yet built, has received intensive study. The
ideas and experience with these units have been incorporated into the design of the
State College reactor, with minor changes as necessary and with the addition of
extra features wherever the inherent safety of the machine or its resistance to
sabotage could be improved.

A tabulation has been included of hazards which could result from operation
or misoperation of the proposed unit. Many safeguards have been included to prevent
injury through such obvious and inherent hazards as exposure to radiation from the
reactor and carelessness in handling radioactive materials in routine operations.

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[page vi]

CONFIDENTIAL

It is demonstrated that pressure rupture or explosion of the reactor by
nuclear reaction is impossible. The "bubble effect" and a large negative tempera-
ture
coefficient prevent any runaway whatever, and insure automatic upper limits to
the level of operation, even if all normal and emergency safety devices fail.

The only important danger inherent in the reactor installation lies in
sabotage through non-nuclear explosion, aimed at wrecking the reactor and blasting
the radioactive solution and attendant vapors and gases over the surrounding area.
Even a catastrophe of this sort would cause little damage beyond that involved in
the catastrophic event itself. A number of precautions against sabotage are in-
cluded
in the design.

CONFIDENTIAL



CONFIDENTIAL

I. THE NUCLEAR REACTOR PROJECT A. GENERAL BACKGROUND

In recognition of the vital role that nuclear phenomena and processes
flow have and increasingly will have in our society, and in discharging her
responsibility to the students of the State and Nation in providing training
opportunities in essential vocational fields, The North Carolina State College
of the University of North Carolina has initiated a training program in
Nuclear Engineering for qualified students. This curriculuma1, which provides
training opportunities at both the undergraduate and graduate level, consists
of (1) classroom instruction in theory and basic information, (2) extensive
laboratory practice of nuclear technology, and (3) research facilities and
opportunities for advanced students.

One of the basic units intended for use as the heart of the advanced
instructional and research programs is a low-power nuclear reactor, of the
uranium "water-boiler" type. It has been proposeda2 that this reactor be built
and operated on the State College campus, by State College, as an unclassified
tool for research and instruction.

In June, 1950, a contracta3 was given by the Atomic Energy Commission
to N. C. State College, directing that the design of a reactor be drawn up,
together with a description of the proposed method of operation and analysis

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[page 2]

of the potential hazards associated with the machine. The report below contains
the items and information stipulated in the contract.

B. ADMINISTRATIVE ARRANGENT 1. Design and Construction Period

It is intended that the Nuclear Reactor be designed, constructed,
and operated by North Carolina State College with the consent, advice and
assistance of the Atomic Energy Commission, the necessary finances being
supplied by State College. The Commission will furnish information and
assistance in the design stage; it may furnish certain instruments, materials
and purchasing facilities during the construction period; and will supply the
fissionable fuel necessary for operation.

At State College, the nuclear engineering training program and the
Reactor Project will be a part of and under the direction of the Physics
Department. During the design and construction phases of the project, the
money necessary for these activities from whatever sources obtained, will be
placed in the State Treasurer's office, along with other college funds, in a
separate account earmarked for the Reactor Project, and made available to the
Physics Department for use in obtaining personnel services, supplies, travel,
etc.

2. Operation

Responsibility for operation of the Nuclear Reactor will also be assigned
the Physics Department. It is intended however that the machine serve as a
tool for instruction and research not for the Physics Department only, but
also for a large group of other departments, organizations and institutions,
many perhaps not even connected with State College. It will be necessary


[page 3]

therefore to arrange schedules of operation, assign priorities to projects,
access costs, etc. No definite rules and regulations for handling these
matters can be made at this time. A few general guiding principles may be
stated, however;


[page 4]

C. CLASSIFICATION

It is the basic intent that the Nuclear Reactor be operated as a
completely unclassified project, on research problems which also are unclassified.
Provisions will be made within safe limits for public observation of the reactor
and its operation. It will be the intent that students and research participants
having access to the reactor and its radiations, but not to classified operation
or construction details of the installations (if any), may participate freely
in the project upon satisfactory evidence to the Scientific Director and the College
Administration of their American loyalty.

Those persons who, because of their intimate connection with the design
and behavior of the reactor must have knowledge of classified information, will be
investigated by the F. B. I. and be given security clearance by normal procedures
followed by the Atomic Energy Commission.

Despite the desire to operate the Reactor as completely unclassified
as possible, it will be the strong intent of State College to guard vigilantly
against the release of classified information to any unauthorized individuals.
The Scientific Director and several of the college staff members associated with
the project will maintain "Q-clearance" status with the Atomic Energy Commission
and will keep up-to-date on declassificable information. In the course of any
investigation, whether considered unclassified or not, should any classified
information be encountered, this will be declared classified and further disclosure
to unauthorized individuals will not be permitted except through regular Commission
channels. Any information of doubtful classification will also be handled as
classified and submitted for clarification of classification status.


[page 5]

D. SAFETY AND SECURITY OF FISSIONABLE FUEL

The enriched U235, in either solid or liquid (sulfate) solution form,
will be delivered to State College at the appropriate time for use in the
reactor. At least three separate containers, of less than 350 g (U235) each,
will be used in transporting the material. The containers will be stored in
fire-proof, combination-lock, storage receptacles, with at least 24" between each
container. In all operations involving fissionable material, no more than 350
grams will be handled at one time or allowed to come nearer than 24" to other
fissionable material. These same rules for handling and storage will be followed
at all subsequent times.

The sulfate solution will be placed into the reactor cylinder through
the sampling tube extending to the top of the concrete shield. (Described later.)
When all the solution has been added, the safe-door covering the end of the
sampling tube will be closed and locked. All beam ports and other openings through
the concrete shielding will be closed by combination-lock, wall-type safe-doors
built into the concrete shielding. Every external opening through which access
to the inside of the concrete shield could be gained will be closed and locked.
Thus, the fissionable fuel will be contained in a closed (stainless steel)
system located completely inside massive concrete shielding in which all external
openings are securely closed.

When the reactor is in operation, or when apparatus is mounted permanently
into one of the beam ports, the safe door on one or more ports will be unlocked
and open. At other times, all openings into the shielding will be securely closed.

In addition, when members of the research staff are not present, the
electrical gear at the control console, to all parts of the reactor, to the crane,
and to other equipment relating to the reactor, will be de-energized and the
switches will be locked.


[page 6]

The reactor building, especially the Reactor Room, and the gates in the
external area fence will be locked, and the area around the building will be well
lighted at night. The college watchmen will direct particular attention to this
building on their rounds of the campus.

The design of the assembly, the precautions listed above and the ex-
tremely
radioactive nature of the material after initial operation, are believed
entirely adequate to insure the physical security from theft of the fissionable
material. No nuclear fuel, other than that in the reactor itself, will be
located at this site. The danger to be guarded against, if any, is in sabotage
aimed at wrecking the reactor or spreading the radioactive fuel around the
neighborhood by non-nuclear (e.g., T.N.T.) explosive blast. This cannot be
accomplished unless the explosive is gotten inside the concrete shield. It is
proposed that the precautions listed above and the design of the reactor will
be adequate to prevent this. The Atomic Energy Commission may, however, prefer
to place on duty at the site a full-time guard or perhaps a guard on duty when
members of the research staff are not present, in addition to the above pre-
cautions
.


[page 7]

CONFIDENTIAL

II. THE NUCLEAR REACTOR A. GENERAL REGIONAL CONDITIONS 1. Location of State College

Raleigh, North Carolina, lies in Wake County, at latitude 35°-47'-20"
north and longitude 78°-40'-40" west. The population is about 65,000.
Raleigh is the capitol of the State, and governmental business engages a
goodly portion of the inhabitants. Meredith College (for women), Peace Junior
College
(for women), St. Mary's Junior College (for women), the State School
for the Blind
, and N. C. State College of the University of North Carolina
are located in the city. Convergence of highways and railroads places
Raleigh in good strategic location for a thriving goods outlet and
commodity distribution center of the State. There are few manufacturing
and heavy industries in the the city.

The campus of North Carolina State College lies toward the western edge
of Raleigh (Figure 1). Hillsboro Street extends westerly from the Capitol
Building at the center of Raleigh, and, at distances between 1 1/2 to 2 5/8
miles from the Capitol, forms the northern boundary of the college campus.
The campus thus extends east and west 1 1/8 miles along the south side of
Hillsboro. The north-south dimension of the campus proper is about 1/2 mile.
The tracks of the Seaboard and Southern railroad run through the middle of
the campus, roughly parallel to Hillsboro Street. Between Hillsboro Street
and the railroad, therefore, is an area of 1000 to 1500 feet in width and
1 1/8 miles in length. Most of the college buildings, except new dormitories
and athletic buildings, lie in this tract. The new dormitories and gymnasia
lie south of the railroad.

There are four general functional groupings of buildings on the campus.
At the eastern end, toward Raleigh, are the College administrative buildings

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[page 8]

and the classroom buildings of the Basic Division. Next westward lie the
classroom buildings and laboratories of the Engineering Division, followed
by the buildings of the Agricultural Division. Finally, on the western end
of the campus, are located the structures belonging to the Textile Division.

2. Topography, Soil Conditions

The land comprising the campus of State College is gently rolling, and
varies between 350 and 420 feet above sea level (Figure 1). "The undisturbed
soil in this area consists primarily of a residual soil developed from
metamorphic rock classified by geologists as Mica Gneiss. This soil is
classified under the Agricultural Soil Survey System as a "Cecil Clay Loam."
It is very plastic, highly impervious, inorganic, red clay-tough near the
plastic limit. In the proximity of Rocky Branch and its tributaries there
occurs alluvial soil of variable nature. The occurrence of natural drainage
ways is limited to Rocky Branch and its tributaries. This stream discharges
into an artificial lake, known as Pullen Lake. The highly impervious nature
of the soil, underlaid with bedrock, precludes the frequent occurrence of
underground drainage ways."b1

3. Meteorological data

Certain meteorological data have been recorded at Raleigh for about 60
years. Certain of these data of interest in the operation schedules of the
reactor are tabulated below:


[page 9]


[page 10]

Inversion layers and other unusual meteorological effects occur, but data
on the frequency and extent of these conditions are not readily available.
Additional data are being obtained.

4. Specific Location or the Reactor

The Nuclear Reactor will be located in one or the other of two proposed
positions on the State College campus. The two locations are shown in
Figure 2.

Position A: South of old Zoology Building. It is planned that the old
Zoology Building shall be removed, but this is not likely to occur in less
than three or five years. With the building removed, this site would
constitute the optimum choice for the location of the Reactor Building.
It is intended that the Reactor facility be located in the space south of
the old Zoology Building and north of the railroad, between the Diesel
Building and the Riddick Laboratories. The facility will be so designed
that (a) suitable interim operation until removal of old Zoology can be
achieved and (b) ultimate expansion, after removal of the Zoology Building,
into an optimum facility can be undertaken.


[page 11]

Should it turn out that the space behind old Zoology is insufficient for
an interim Reactor facility, or should further study reveal a strong un-
desirability
from the safety standpoint of crowding the Reactor facility
for interim operation into cramped space immediately adjacent to a classroom
building, then the Reactor facility will be located in Position B.

Position B: North of Old Zoology Building. The Reactor will be located
between Daniels and Polk Halls, north of University Drive, if the space
south of old Zoology proves unsuitable. If the Reactor is placed in this
location, a three or four story new physics laboratory building will also
be placed in this area in order to reestablish an architecturally acceptable
landscape plan for this portion of the campus. The new building, as shown
on Figure 2, would extend in an east-west position between Polk and Daniels
Halls, in line with the north ends of these two buildings. Part of the
Reactor laboratories would be located in this new building.

In either of these two locations the environmental conditions are
essentially the same. Six large buildings are nearby: Withers, Daniels,
Riddick, Diesel-Mechanical, Polk, and, of course, Old Zoology. Patterson
Hall is directly north, 1020 feet from Position A and 980 feet from Position B.
The elevation is 410 to 415 feet above sea level: one of the highest
elevations in the entire neighborhood. The general slope is south, east
and west from these locations, 2000 feet to the east and 1500 feet to the west
respectively are located the troughs of north-south surface drainage ways.
In the west trough there is also a small underground water channel. There
are no other known surface or underground flow channels in the vicinity. The
drainage ways from this area lead eventually to Rocky Branch.

The non-college activity nearest the proposed reactor location is a
small shopping center on Hillsboro Street, 2200 feet north. East, south and
west the college property extends 3000, 6000, and 3000 feet respectively from


[page 12]

the reactor site. The nearest college dormitories are 2300 feet southwest;
these are at 40 feet lower elevation than the reactor site. The Seaboard
and Southern Railroad
runs east and west at 500 and 1200 feet south re-
spectively
of the two proposed reactor sites. The railroad elevation is
30 feet lower than that of the reactor site.

B. BUILDING

A building is being designed to house the Reactor and the activities
associated therewith. Figure 3 shows sketches of the functional components of the
proposed building. The finished design may have somewhat different floor plan
from that shown, but it is intended that it contain the four essential components
shown:


[page 13]


[page 14]

C. VENTILATION SEWAGE AND WASTE DISPOSAL 1. Ventilation

Large blowers located high above the reactor in the Reactor Room will
draw air via ducts from all areas of the building through filters, and
discharge it to the building stack which will in turn discharge the air into
the atmosphere 25 feet above the roof of the highest buildings in the nearby
area. Within the building, therefore, air will flow from areas of low
radioactivity toward the area of highest activity. It is not intended that
the stack will be used for routine disposal of radioactive gases into the
atmosphere. These will normally be handled in other ways as described
below. Traces of activity or occasional inadvertent small gas releases may
accompany or contaminate the ventilating stream of the air through the
blowers to the atmosphere. The stack discharge is placed high in the atmos-
phere
to provide safe disposal of these traces and infrequent inadvertent
releases. Careful and continuous monitoring of the ventilating stream will
be maintained. The source of any increase in activity will be promptly
located and handled by prescribed methods.

2. Sewage

Radioactive liquids and solids will not be permitted to enter the city
sewage system. Three separate sewage systems will be provided in the
building. (a) Waste from the toilets, storm sewers, drinking fountains,
etc., will go directly into the city sewage system. (b) Drain lines from
the laboratories or any area in which radioactive materials are handled will
go to one of two underground (but accessible) holding tanks, and from there
to the city system. Radioactive solids or liquids will not normally be
placed in these drain lines. It is possible, however, because of their
proximity to working areas, that accidental spills or disposal of radioactive


[page 15]

materials may find their way into these lines. Therefore, continuous
monitoring at the entrance to the holding tank will be maintained; and should
appreciable quantities of radioactive material inadvertently reach the tank
from the laboratories in the building, it will be held there and handled by
prescribed methods, and then be allowed to enter the city sewage system only
under predetermined safe conditions. Meanwhile the drainage from the
laboratories will be switched to the second holding tank. (c) Short-lived
radioactive materials may be dispatched down one or two special drains to an
underground (but accessible) holding tank, which has no connection with the
city sewage system. When the activity has decayed sufficiently, the material,
by positive effort (not automatically), may be disposed of as non-active
waste.

Long-lived materials of solid or liquid nature will be accumulated in
closed, shielded, non-breakable containers designed for safe subsequent
transportation to Oak Ridge, Tennessee for disposal. Radioactive gases
will be accumulated in non-breakable cylinders, or absorbed in liquid or
solid absorbents, and handled as above.

Thus, no activity whatever above trace level will enter the city sewage
system, or be dispersed into the atmosphere or into the ground.


[page 16]

D. REACTOR DESIGN 1. General

The reactor assembly consists of the reactor itself which is a stainless
steel cylinder containing a uranyl sulfate solution, the graphite reflector,
a limonite or barytes concrete-lead shield, a graphite thermal column, sample
exposure-neutron beam ports, control and safety rods, cooling coils, and
auxiliary apparatus and systems. Each of the major component parts will be
described separately below. The overall design is intended to produce a
simple and relatively inexpensive instrument that will provide maximum useful-
ness
and flexibility, and minimum potential hazards. The water-boiler type
of reactor was chosen because of its simplicity, its proven usefulness, its
inherent safety characteristics, and the long experience of satisfactory
operation of the Los Alamos unit. With this type of reactor, a relatively
high radiation flux can be obtained with simple construction and equipment.
The two chief disadvantages of the liquid-type of reactor are (1) the mobility
of the fuel creates a design problem in avoiding loss if leaks should occur,
(though the mobility is also an advantage in that fuel transport, handling,
etc. by remote control, through pipes, pneumatic pressure, etc., is
relatively easy,) and (2) decomposition by radiation of the fuel solution
into gaseous products creates problems of gas disposal and solution replenish-
ment
.

Since this reactor will be operated in the relatively populous environs
of a college campus and as an unclassified establishment relatively accessible
to the public, an attempt has been made to incorporate into the design
(1) exceptionally large safety factors against potential hazards and
(2) unusually extensive precautions against sabotage.


[page 17]

2. Shielding Arrangement

The major volume of the reactor assembly consists of concrete shielding.
The shielding is designed to perform the primary function of attenuating to
a safe level all radiations from the reactor and its accessories. It is
intended that a "heavy" concrete will be used; i.e. one in which the usual
crushed stone or gravel aggregate is replaced with a metallic ore which
possesses greater radiation attenuating ability. Limonite (iron ore) or
Barytes (Barium Sulfate-Iron Oxide) aggregates are being considered.

A lead shield (described later) will surround the reactor inside the
concrete shielding. Using shielding data empirically obtained at Oak Ridge
and Los Alamos and calculating the attenuation of radiations expected from
the reactor, it is found that a total of 30 inches is sufficient to reduce
the reactor radiation to a safe level. However, because of various pieces
of apparatus to be imbedded in the shield, and to provide an adequate margin
of safety, the concrete around the reactor will be made 6 feet thick.

The concrete in this particular assembly performs several secondary
functions in addition to the primary function of attenuation; (1) It provides
security of the fissionable material against theft. Hence, the design must
minimize accessibility to the interior in which all parts of the reactor
system are enclosed. (2) It is the primary defense against possible attempts
to sabotage the installation, hence, no openings may be left into which
explosives or other damaging materials could be placed. (3) It houses the
auxiliary systems of the reactor (exposure ports, control, cooling, gas
disposal, etc.) and these much be reasonably accessible in proportion to
their demand for use or maintenance.

Numerous channels and openings are provided in the concrete shield.
The entire shield is octagonal in shape, 17 feet in diameter and 9 1/2 feet
high. The assembly is composed of an octagonal poured concrete base, 18 inches


[page 18]

thick and 17 feet across and two poured concrete super-structure sections
comprising three octagonal faces in each on opposite sides of the base. The
super-structure extends 96 inches above the base slab, (Figure 4). Between
the two poured concrete super-structures a large open channel is left. This
opening is 5 feet wide and 92 inches deep and extends from one octagonal
face across to the other. Inside this volume will be placed the reactor,
the reflector, the gas handling system, the instrumentation-control mechanism,
and the graphite thermal column. The remaining volume will be filled with
slabs and blocks of concrete. Various holes, ports, and trenches are pro-
vided
in the poured concrete sections to accomodate various auxiliary apparatus
of the reactor assembly.

When the reactor assembly is complete, the concrete shielding will be
so arranged that no access whatever to the interior can be accomplished ex-
cept
(1) through beam-ports which are closed when not in use by combination-
lock
safe doors, or (2) by removal of external slabs of concrete, which weigh
not less than 5 or 10 thousand pounds each. The concrete slabs, in addition
to their weight, are interlocked into the assembly in such fashion that
their removal is impossible unless the "cap-slab" on top of the assembly
is removed first. The "cap-slab", in turn, is securely bolted in position
and the bolts are secured against removal by locks. A completed assembly
of shielding components is shown in Figure 5.

3. The Fuel Container.

Figure 6 shows the fuel container and its attached components. Stainless
steel (18 - 8), type 347 is used throughout the fuel system. and in all struct-
ural
parts in contact with the solution or its vapors. Thickness of the
reactor walls is 1/16 inch. The fuel container is a cylinder of 27.2 cm.
diameter and 27 cm. height. (See Nuclear Fuel, below.) There are no


[page 19]

openings in the bottom of the cylinder or on the vertical sides. There
are 13 openings in the top surface of the reactor cylinder.

All re-entrant and connecting tubes are welded into the reactor. All
joints and connections in the reactor are welded. The entire assembly is
built to withstand an internal pressure of 100 p.s.i.a. without rupture


[page 20]

and without leaks detectable with a helium mass-spectrometer leak-detector.
The reactor must meet these specifications by actual test before initial use.
All tubes connecting to the reactor project vertically upward for 18 inches.
At this point, all except the exposure-tube, the control rod sheaths and
the level indicator, turn through 80° or 90° to an approximately horizontal
position. Beyond the bond or elbow, a leak-proof coupling is provided in each tube.

4. The Cooling System.

The heat generated in the reactor during operation must be removed.
The cooling system designed to accomplish this consists of three helical
coils of 1/4" stainless steel tubing of 30 feet total length inside the
reactor cylinder. At maximum anticipated power level (10 K.W.), one
gallon per minute of water from the city mains through each of the three
coils, or a total of 3 gallons/minute will dissipate the heat. Since the
coils are immersed in the corrosive uranyl sulfate solution, they must be
made of stainless steel. Water comes from the city system to the three
coils through separate automatic pressure reducing-regulating valves. A
thermocouple indicates the water temperature. The pressure regulating
controls for the valves are mounted on the Control Console. From the
three valves, the water goes directly to the three coils inside the reactor.
Thermocouples indicate and record the water temperature down stream of each
coil. The exit lines from the coils join into one line at a point about
1 foot from the reactor, and through the single 3/4" tube the water flows
downward inside the concrete shield to an underground pipe trench and along
this trench to an underground holding reservoir outside the reactor building.

A small amount of short-lived activity is built up in the water during
its passage through the reactor.


[page 21]

Data supplied by the City Public Utilities Department on water analysis
made over a period of years indicate that the average mineral content
of Raleigh city water is:

SiO29.8 p.p.m.S0411 p.p.m.
Fe0.02Cl4.9
Ca8.7F0.1
Mg1.4NO30.1
NatK4.2HCO314.0
Mn0.0CO33.9

The number of atoms per cc of water of each of these chief constituent
elements and the nuclear characteristics of the particular isotope of interest
are indicated below. The calculations are based on a 2 second exposure in a
thermal neutron flux of 1012.

IsotopeAtoms/cc of H20Slow neutron
cross section x 1024
Resultant half
life activity
Resultant Activ-
ity
Disinte-
grations
/cc/sec.
F193.4 X 10150.0112s3.35
Si309.8 x 10150.122.8 hrneg
Fe544.0 x 10132.14 yrneg
Fe582.0 x 10120.3246 dneg
Ca445.8 x 10150.4315 dneg
Mg265.3 x 10154.810 m625.
Na231.4 x 10170.514.8 hneg
S241.5 x 10160.2687 dneg
Cl351.2 x 101753.02 x 106 yneg
Cl374.1 x 10160.637 m9.5
0187.5 x 10190.0002231 s110

Actually, the exposure of each cc of water should be less than 2 seconds,
at normal cooling water flows of 1 gallon per minute through the 1/4" cooling
pipe, and the average thermal neutron flux will probably be less than 1012.


[page 22]

The induced activity is therefore probably somewhat exaggerated. Even so,
the resultant activity is less than 1000 disintegrations per second in each
cubic centimeter of water. Much of this radiation is internally absorbed
in the water.

Assuming no shielding and no internal absorption of radiation by the
water, 10 gallons of freshly irradiated water would produce a radiation
dosage at the rate of 0.08R/8hr. to a person standing 5 feet away. After
one hour, this dosage rate reduces to 0.0008R/8hr. at 5 feet.

The water therefore is sent to a holding tank where the activity is
allowed to die away before discharge into the sewers. The 1800 gallon holding
tank is large enough to hold the water form 10 hours of maximum level operation.

Radiation detectors, with indicating meters on the Control Console, are
placed at the entrance and exit of the holding tank. If any increase in
activity should occur, the cause will be determined and escape of active
material from the tank to the sewer will be prevented.

If a leak in the cooling coils inside the reactor should occur, the
water will tend to flow into the reactor solution instead of the converse,
because of the pressure in the coils. Should a leak occur, and sufficient
water enter the reactor for the liquid level to rise, an interlock on the
level indicator will operate and shut down the reactor and close the valves
in the water line to the cooling coils.

5. The Nuclear Fuel.

The water boiler at Los Alamos operated first with uranyl sulfate
solution and later with uranyl nitrate solution. The change was made
primarily because it was expected that periodic chemical clean-up of the
solution would be necessary, and the nitrate seemed much more amenable to
this operation than the sulfate. After extended operation it has become


[page 23]

apparent that necessity for clean-up processing will be an extremely rare
occurance, hence this reason for choice of nitrate is of little importance.

The nitrate perhaps does have a lesser tendency to corrode than does
the sulfate, but experiments at Los Alamos and Oak Ridge indicate that
sulfate corrosion of stainless steel is extremely small at temperatures
below 100°C.

The solubility of uranyl nitrate is loss than that of the sulfate. Once
dissolved however there appears no difficulty in preventing precipitation of
either the nitrate or the sulfate, provided the solution is kept sufficiently
acid.

There are three chief advantages in using the sulfate solution: (1) the
boiling point is somewhat higher for the sulfate solution than for
the nitrate solution, (2) the radiation decomposition into gaseous
products is much lower, is about half in fact, that of the nitrate, and
(3) the neutron absorption is less in the sulfate than that in the nitrate.

It is intended, therefore, that the reactor described herein will operate
with a solution of uranyl sulfate as the nuclear fuel. Uranium highly en-
riched
in U235 will be used. The amount of uranium needed and the dimensions
of the cylindrical reactor are determined approximately by comparison with the
amount and dimensions in the spherical Los Alamos Reactor, with appropriate
conversion factors. A detailed, accurate evaluation of amounts and dimensions
will be made later from considerations of reactor theory and nuclear constants
when exact arrangements of reactor, reflector, etc. have been decided.

The Los Alamos Reactor contains 12,600 cc of uranyl nitrate in which
839 grams of U235 are dissolved. 764 grams are needed to produce criticality
at 20°C, and the extra 75 grams are needed to overcome the negative temperature
coefficients and provide a useful excess reactivity. A cylindrical reactor,
because of its larger surface:volume ratio, requires 1.14 times as much


[page 24]

material as a spherical reactor, other conditions being equal. On this basis,
the cylindrical reactor solution should have a volume of 12,600 x 1.14 =
14,360 cc. Using optimum height to radius dimensions for a cylinder of
H/R 1.848, gives height 24.8 cm, and diameter = 27.2 cm. A cylinder of
27.2 cm diameter and 27.0 height would permit 2.2 cm depth of unoccupied
volume in the top of the cylinder. This depth becomes 2.9 cm when the sulfate
Solution and the difference in cooling coils is taken into consideration.

About 840 grams of U235 in the cylindrical reactor are required to
produce criticality at 20°C, and about 60 grams more, or 900 grams, at 80°C.
The negative temperature coefficient is such that the critical mass increases
about 0.9 grams per degree centigrade. Fifteen additional grams are added
to insure sufficient excess reactivity for useful experimentation. The total
U235 required, therefore, will be about 915 grams.

6. The Reactor Envelope

The stainless steel reactor itself is surrounded by a second envelope
made chiefly of aluminum (Figure 7). The 27.5 cm O.D, (diameter) cylindrical
reactor is enclosed in a 28.3 cm I.D. cylinder which is closed at the
bottom, underneath the reactor, and which extends upward 16" to a flanged
connection in the lower surface of a much larger aluminum cylinder. The
lower part of the smaller cylinder is made of stainless steel, so that it will
not be chemically attacked immediately in case a leak should occur. The
reactor itself rests on and is supported by the lower end of the Reactor
Envelope. The larger upper cylinder, 48" in diameter and 42" high, flange-
closed
at top and bottom, together with the connecting smaller cylinder
fitted around the reactor, constitute the Reactor Envelope.

The purposes of the Reactor Envelope are (1) to catch any liquid which
should inadvertently leak from the reactor; (2) to retain for leisurely


[page 25]

disposal any radioactive gases which inadvertently escape from the reactor;
and (3) to serve as a safety volume into which the reactor contents could
expand without wide liquid or gas dispersal in case of rupture of the
reactor system. This envelope-chamber is not absolutely vacuum-tight and
does contain several imperfectly fitted joints through which gases under
pressure inside the volume could slowly escape. The envelope, however,
would reduce any leak to a very low rate so that the ventilation system
could dispose of the escaping gases without permitting the room to become
contaminated. The gases inside the envelope can be pumped out slowly through
purge lines to absorption traps or pumped directly to the stack for disposal.

The volume of the envelope (exclusive of the space occupied by motors,
etc.,) is about 40 cubic feet, or about 45 times the total volume of the
reactor itself, hence any pressure inside the unfilled volume of the reactor
should be tremendously reduced if it should expand into the envelope volume.
The reflector around the reactor will be so arranged that the line of least
resistance for a pressure release around the reactor. will be upward into the
large volume of the envelope. That is, the 16" of graphite on top of the
reactor, inside the reactor envelope is loosely packed powder which would be
pushed aside by a pressure release below.

An atmosphere of inert gas under slight positive pressure will
normally be maintained in the envelope. Periodic sampling will quickly
reveal the pressure of any gases leaking from the reactor. The positive
pressure will retard the escape of gases from the reactor through any leak
which should occur.

The water lines to the cooling coils, the refueling sampling line
to the reactor, the gas disposal tube, etc. from the reactor penetrate
the wall of the reactor envelope through screwed-in, spring-tightened gas


[page 26]

seals. These seals are not vacuum-tight, but will only permit the escape of
negligible amounts of gas unless the pressure inside becomes extremely
large. These tubes are all arranged with union couplings in the lower
4" of the large cylinder of the reactor envelope. If these unions are
uncoupled, and the lower flange seal of the large cylinder is broken (by
removal of screws), the upper part of the reactor envelope, with the un-
connected
tubes and pipes, can be removed. This would only be necessary
under certain emergency conditions, described later.

The reactor is not concentrically placed inside the slightly larger
cylinder of its enclosing envelope. The reactor touches one side of its
envelope, which leaves a gap of about 0.8 cm on the opposite side between
the reactor and the envelope. In this space are located (1) a thin strip of
cadmium which moves vertically in a guiding scabbard and serves as a shim-
control
rod, and (2) a small tube extending downward to the bottom of the
reactor envelope, to the lower surface of the reactor, which provides a means
of removing liquid from the envelope in case a leak in the reactor develops.
The remaining space is filled with tightly packed graphite powder.

7. The Reflector.

The reactor envelope, immediately underneath the reactor, rests upon a
16" thickness of graphite blocks, and is also surrounded on the sides con-
tiguous
to the reactor by 16" of graphite blocks.

The reflector is placed around the reactor (a) to decrease the amount of
U235 needed in the reactor and (b) to increase the value of the radiation
flux at the surface of the reactor. Highly purified graphite, shaped from
4" x 4" rectangular bars to fit snugly against the reactor surface is used as
the reflector. The graphite is placed both inside and outside the reactor
envelope, so that a thickness of at least 16" is present on all surfaces of
the reactor.


[page 27]

Inside the reflector envelope, above the reactor and in the narrow
channel around the sides, powdered graphite to a depth of 16" serves as part
of the reflector. This powder is packed sufficiently to prevent any
effective shifting of the reflector during reactor operation.

8. The Second Liquid Catch Basin.

If a leak should occur in the reactor, the liquid will be caught in the
bottom of the reactor envelope. Any liquid in the reactor envelope would be
in close contact with the reactor, and part of the same nuclear fuel
accumulation. The control and safety rods of the reactor, therefore, would
serve to prevent inadvertent nuclear reaction because of accumulation of
the leaked liquid.

In case a leak or rupture should occur in turn in the reactor envelope,
a second catch basin for liquid is placed underneath the first. The secondary
catch basin has an upper and a lower part. The upper part is an aluminum
cylinder, open at the top, into which the lower end of the reactor envelope
loosely fits. The cylinder is filled to within 2" of its upper end with
snugly packed graphite blocks. The reactor envelope rests solidly on this
graphite for support. Small channels through the graphite permit any liquid
leaking from the reactor envelope to trickle down through the graphite to
the lower part of the catch basin. Calculations show that a nuclear chain
reacting condition in the graphite below the reactor will not be closely
approached, even if all the solution from the reactor is thoroughly and
uniformly inpregnated throughout the graphite, which is not likely to occur.

The lower part of the catch basin consists of a broad shallow chamber
in which the liquid from above may be caught. There are no control rods in
this vicinity, hence the unfavorable geometry of the flat catch basin is
necessary in order to prevent uncontrolled nuclear reactions in case all of
the nuclear fuel should leak down to this location.


[page 28]

9. The Lead Shield.

An aluminum cylinder, 52 inches (O.D.) in diameter and 48" high, open
at the top, but lined on the sides and bottom with 2 inches of lead, encloses
the reactor. Holes are cut in the sides for the passage of exposure parts
to the reactor. An additional layer of lead, unattached to the cylinder,
is placed externally around the sides of the cylinder. The purpose of this
heavy metal layer is two-fold: (1) The metal acts as an attenuator for the
gamma radiation from the reactor, thus reducing the amount of concrete shield-
ing
needed; and (2) in case the reactor must be removed from the assembly
while it is highly radioactive from undecayed fission products, it may be
removed inside the metal lead-lined cylinder, which then serves as a
shielding container for the transportation or storage of the reactor. The
"port holes" may be easily filled and a cover may be placed on the top to
provide a complete shield after the upper part of the reactor envelope is
removed.

The cylinder serves incidentally as a tertiary catch basin for leaked
solution, in case the first and second catch basins fail.

10. Safety and Control Rods.

The reactor is provided with two identical boron rods and one cadmium
"rod". One of the boron rods serves as a safety rod and the other as a
control rod. The cadmium "rod", a 0.02" thick, 2" wide strip of cadmium
mounted flat against the outside vertical wall of the reactor, serves as a
shim control rod. The boron rods each consist of 8" of enriched (B10)
sintered boron ([rho] = 1.5 - 1.7) inside of a 5/8 o.d. thin walled stainless
steel tube. The boron tubes are mounted vertically inside of stainless steel
scabbards which are re-entrant through the top surface of the reactor, 8" down
into the reactor. The boron rods are located 4" from the central vertical
axis of the reactor and approximately 100 radial degrees from each other.


[page 29]

In these positions the boron rods are each "worth" about 80 grams of
U235 in the solution. Thus either rod alone is equal to the total excess
U235 in the solution above that required for criticality at room temperature.
The shim rod is "worth" about 10 or 12 grams.

The stainless steel tubes containing the boron in their lower ends
extend upward inside the (re-entrant) scabbard tubes about 18" to an
electromagnet connection to vertical, motor driven "rack and pinion" rods.
The "rack and pinion" rods may be raised or lowered by signals to their
respective motors from the Control Console. The boron rods, likewise,
through the electromagnet connection, are raised and lowered with the "rack
and pinion" rods. Should current to the electromagnet be interrupted the
connection is broken and the boron rod drops 8" into the reactor inside the
re-entrant scabbard tube. The shim rod is likewise raised and lowered by
motor drive.

In normal operation, one boron rod (safety) is hoisted completely out of
the reactor to a poised position from which it can drop back into the reactor.
The shim rod is partially removed, and in that position oscillates up and down
in response to electronically amplified signals which attempt to move the
shim rod to counteract fluctuations in the operating level of the reactor.
The other boron rod (control) is partially withdrawn to such level as will
cause the reactor to operate at the desired level.

11. Instrumentation and Control.

The instrumentation included in the reactor installation is intended:
(1) to provide the operators with knowledge of all relevant conditions and
processes occurring in the assembly and to record such information when
desirable; (2) to furnish the operators with means of guiding, controlling,
and regulating all processes.


[page 30]

The chief components of the instrumentation system are: (1) the sensi-
tive
devices: thermocouples, pressure guages, radiation
detectors, etc., in and around the reactor assembly; (2) the signal transmitting system: cables,
pressure leads, etc.; (3) the indicating and recording mechanisms at the
operator's location; and (4) the controlling, regulating devices at the
operator's location.

A control console will be located outside the reactor room in such a
position that an operator at the console can view the inside of the reactor
room through a large water window, 16 inches thick, (as protection from stray
radiation). On the console will be located the indicating, recording, and con-
trol-regulating
mechanisms. Trenches under removable sections of the floor,
loading from underneath the reactor to a point underneath the console, will
provide the location of the signal transmitting devices.

The measurement and control of the operating level of the reactor is by
far the most important component of the instrumentation system. Temperatures,
pressures, water flow, gas disposal, monitoring data, etc., are also of vital
importance. The vaious important components of the instrumentation system
are described below:


[page 31]


[page 32]


[page 33]


[page 34]


[page 35]


[page 36]

12. Operating Level; Radiation Fluxes; Reactivity of Solution.

All component parts of the reactor involved in determining the power
level of operation (shielding, cooling coils) have been designed to permit
steady state operation at 10 Kilowatts. It is probably that, in actual
operation, power levels in excess of 5 KW may not be desired for a long
time. Indeed, a great deal of work will be performed at 1 KW or less.

Table 1 lists the estimated and calculated radiation fluxes at various
points in and about the reactor at 10 KW power output.

TABLE 1. VARIOUS FLUX DENSITIES AT VARIOUS POSITIONS (10KW)

1. Surface of Reactor Vessel.

From experimental work on a reactor similar to the one described in this
report estimates of radiation fluxes on the surface of the reactor are:

Gamma rays5.4 x 1011[gamma]'s/cm²sec.2Mev.
Neutrons fast1 x 1011n/cm²sec.
Neutrons slow3 x 1011n/cm²sec.

2. Re-entrant exposure tube inside the reactor.

Based on the neutron distribution in the cylindrical reactor the above
figures would require the following values at the center of the cylinder
which we will consider as the values in the re-entrant exposure tube.

Gamma rays5.4 x 1011[gamma]'s/cm²sec.2Mev.
Neutrons fast5.5 x 1011n/cm²sec.
Neutrons slow1.5 x 1011n/cm²sec.


[page 37]

3. Experimental Port.

At the external end of port the flux can be estimated by assuming only in-
verse
square law acting. This will be approximately true even for [gamma] rays.
On this basis, the calculated values are:

Gamma rays2.0 x 109[gamma]'s/cm²sec.2Mev.
Neutrons fast2.0 x 108n/cm²sec.
Neutrons slow1.2 x 109n/cm²sec.

4. Thermal Column.

At external surface of column (5 ft from reactor) the slow neutron flux will
be 3 x 107 n/cm²sec. with about 5.2 x 10² fast neutrons, (i.e., 60,000:1).
There will be [gamma] rays which will come from the absorption of slow
neutrons when Cd sheet is in place to absorb slow neutrons.

Gamma raysfrom Cd(n,[gamma])when Cd in place
Neutrons slow3 x 107n/cm²sec.
Neutrons fast5 x 10²n/cm²sec.

5. Along Thermal Column.

Tabulated below are the neutron flux available at points along column:

Distance from
Lead Shield
1 foot2 feet3 feet
n(slow)5 x 1010 n/cm²sec.1.2 x 1010 n/cm²sec.3.5 x 108
n(fast)4 x 106 n/cm²sec.5 x 105 n/cm²sec.1 x 104

As the reactor is operated, the fuel solution and surrounding materials
become radioactive. The total radiation from the reactor derives from three
sources: (1) fission of uranium, which instantaneously releases neutrons and
gammas, with relatively fewer alpha and beta particles; (2) fission products,
which are highly radioactive when first formed, and they release a few
"delayed" neutrons and many beta and gamma particles; and (3) radioactive


[page 38]

materials artificially (activated by the radiation from (1) and (2). This
induced activity consists chiefly of betas and gammas. When the reactor is
not in chain reacting condition, radiation is not produced by (1) fission,
but does continue from (2) and (3): fission products and induced activities.
Both these latter materials decay in activity with half lives characteristic
of the particular isotopes involved. The composite total of the half lives
involved results in a characteristic decay pattern for the reactor contents.
The "composite" half life has been found to be about 55 seconds.

When the reactor is brought to a chain reacting condition at a certain
fissioning rate, after a period of inactivity during which the previousl
induced activity decayed to a low level, radiation from the fissioning atoms
is immediately produced and fission products begin to accumulate. Induced
activity in the surrounding materials also begins to build up. assuming that
the fissions continue at a constant rate, the total radiation steadily in-
creases
due to the contribution of the fission products and the induced
activities. The increase continues until the decay of the non-fission
sources is equal to the rate of formation. This will require a very long
time, for a small fraction of the fission products are very long-lived.
Thus, a "steady state" condition involves a small but gradual rise in total
radiation from the reactor, even though the fission rate remains constant.

Table 2 below contains calculated values of the total activity of the
fuel solution at various elapsed periods after shut down from various
operating levels. It is assumed that the gaseous fission products escape
from the reactor as they are formed.


[page 39]

TABLE 2. ACTIVITY IN CURIES OF FUEL SOLUTION

Operation
Time (days)
PowerTime in Days After Shut Down
0.0010.1151520
110KW2.1x1044.3x10³3.1x10²1.8x10²0.52x10²0.38x10²
10102.4x1046.7x10³2.6x10³1.0x10³4.0x10²3.0x10²
100101.8x1051.0x1044.3x10³2.3x10³1.4x10³1.2x10³
112.1x10³4.3x10²0.31x10²0.18x10²5.23.8
1011.8x1041.0x10³4.3x10²2.3x10²1.4x10²1.2x10²

13. The Thermal Column.

A large portion of the neutrons emerging from the surface of the reactor
are "fast" neutrons, i.e., their energies and velocities are high. A great
deal of interesting research, however, involves the use of "thermal," or slow
neutrons. Fast neutrons may be slowed by collision with light, low neutron-
absorbing
atoms, of which carbon is an excellent example.

The neutrons from one side of the reactor, therefore, are allowed to
penetrate several feet of graphite, highly purified to remove neutron absorbing
contaminants and, as they emerge, a large percentage have velocities in the
thermal region.

The thermal column of graphite is shown in cross-section in Figure 8.
Four exposure ports into the graphite are provided. Table 1 lists the
anticipated values of the radiation flux at various positions in the graphite

14. Sample Exposure Ports.

It is anticipated that several sample ports may be used simultaneously
for exposure of samples to the radiation from the reactor. Also, it may be
desirable to "tie-up" permanently the bean from one or more ports with a


[page 40]

large piece of special apparatus. Hence provision is made for a large number
of exposure-bean ports through the shielding into the region of the reactor,
though only one or a few of these may be in use at any given time.

There are altogether 12 exposure ports. (Figure 9 )

Seven extend horizontally from the outside surface of the concrete shield
inward to the surface of the reactor envelope, along extended diameters,
respectively, of the reactor, of these seven, one runs along the horizontal
axis of the thermal column.

Four exposure ports extend entirely through the assembly, from the out-
side
surface on one side through the interior of the assembly, to the outside
of the shield on an opposite side. Three of those four traverse the thermal
column, perpendicularly to the horizontal asix of the column. The fourth
is tangent to the surface of the reactor envelope.

The twelfth exposure port extends vertically downward through the top
surface of the concrete shield to the top surface of the reactor itself as a
one inch tube, which continues downward into the reactor as a re-entrant tube
to a depth of 8". Small samples in this tube are exposed to the highest
possible radiation flux.

In order to achieve economy and convenience in construction, all exposure
ports are standardized to a single pattern (except the 1" vertical port).

A 3.5" i.d. metal tube in the concrete is placed in position and permanently
secured there by surrounding it with the poured concrete of the reactor
shielding. Subsequently, the 3" i.d. "lining tube" is inserted into the
first tube so that it provides a continuous passage from the outside of the
shielding to the surface of the reactor envelope. The "lining tube" can be
removed should it interfere with repair work around the reactor inside the
concrete shield.


[page 41]

The external end of each exposure port terminates in a heavy "burglar-
proof
", combination-locked safe door. When not in use, the beam ports are
plugged by inserting successively smaller telescoping tubes inside the "lining
tube" until the passage is closed. The safe door is then closed and locked.

All exposure-ports emerge horizontally from the shielding at a level of
24" above floor level. The emergent beams thus would strike a person who
carelessly stepped into the beam path on the legs, rather than in a more
vital region. Radiation beams from the ports traverse paths across the
Reactor Room 24" from the floor level, to respective openings in the building
wall which lead to underground radiation traps outside the building.

To permit usage of lager apparatus than could be accomodated at the
24" level from the floor, trenches four feet wide and two feet deep, ex-
tending
from the respective faces of the reactor to the wall of the room,
are provided. These trenches are normally covered when not in use by
movable sections of the floor of the room.

15. Sampling-Replenishing Lines.

For the purpose of adding solution to the reactor or withdrawing
solution (samples, or complete removal) a Sampling Line is included in the
reactor design. (Figure 7). A 3/8" stainless steel tube, re-entrant into
the reactor, extends from. the inside bottom of the reactor upward through a
coupling in the wall of the reactor envelope, then at a slight incline from
the horizontal, through a submerged trench in the concrete, to terminate at
a cutoff valve just underneath a combination-lock safe-door in the top
surface of the concrete shield.

By unlocking and opening the safe door, the end of the sampling tube is
exposed. Addition of fluid can be readily accomplished by gravity flow into
the reactor. Also, by simple connection to a vacuum pump, preceded by a
liquid trap, solution may be quickly removed from the reactor.


[page 42]

A second tube, the Liquid Salvage Line, lies closely beside the first,
but extends to the bottom of the reactor inside the reactor envelope enclosing
the reactor. With this tube, liquid can be removed from the reactor envelope
in case of leaks, etc.

16. Solution Level Indicator.

For measuring the level of liquid in the reactor two methods amy be used:
(1) The pressure required to bubble air (or other gas) backwards through the
sampling lines to the bottom of the reactor can be measured and, knowing the
specific gravity of the liquid, the solution level can be calculated. (2) A
liquid level indicator is provided for accurate level measurement when the
reactor is nearly full. The latter instrument is described below.

A welder or stainless steel tube of 3/8" inside diameter projects 24"
above the top surface of the reactor (Figure 10). At this point the 3/8" tube
is flange sealed with an insulator gasket to a 2" x 3" stainless steel sylphon
bellows. From the movable top end of the sylphon bellows a 1/8" steel rod
projects 25 inches downward to make contact, on its sharpened point, with
the liquid in the reactor. The sylphon and its attached contactor, being in-
sulated
by the gasket from metal contact with the reactor, may be made the
anode of a low voltage electrical circuit of which the reactor and its electro-
lyte
liquid is the grounded portion. The sylphon bellows may be compressed
by means of pneumatic pressure inside a sealed-on metal chamber enclosing the
sylphon. When pressure is exerted, the sylphon is compressed and the pointer
lowers to make contact with the liquid. When this occurs, an electrical
signal appears on the operators control console. The position of the pointed
is calibrated against the pneumatic pressure. The distance of vertical travel
is three inches. When the liquid is within 3 inches of the top surface of the
reactor, therefore, its position may be determined with high accuracy.


[page 43]

17. Gas Disposal.

When the reactor is in operation, a very small volume of gaseous fission
products will be released. These gases result from the fission of uranium
into elements of gaseous nature near the middle of the periodic table. Most
of the gases will be highly radioactive, but most of the radioactivity will
be quite short lived.

Estimates have been made of the gaseous fission products expected
from the reactor, calculations of the total radioactivity expected, and
of the decay of the radioactivity Figure 11).b1 About 2.5 x 1014 disinte-
grations
per second, or 7000 curries, occur initially from the atoms of the
fission product gases produced per kilowatt minute. After 4 hours, however,
the activity is 5.0 x 109 disintegrations per second, or 0.15 curios from a
kilowatt minute of fission product gas, a decrease in activity by a factor
of over 50,000 in 4 hours.

The fission product gases will be accompanied by much larger volumes of
other gases resulting from the radiation decomposition of the water molecules
in the fuel solution into hydrogen and oxygen. The hydrogen and oxygen will
have negligible radioactivity, but these gases do constitute a mixture of
explosive proportions.

The "gas problem" for the reactor in maximum normal operation (5 KW)
consists therefore in disposal of 40 liters per hour of a hydrogen-oxygen
mixture in which a trace of highly radioactive fission products gases are
intermixed. There are several possible means of disposing of these gases:


[page 44]


[page 45]

It is quite certain that some method or combination of methods can
be devised to provide a completely satisfactory system of handling the
gases. No dispersal to the atmosphere will be permitted until the
activity is decayed sufficiently to be harmless.


[page 46]

If the system were to be built at the time of the present writing,
the combination of methods described below would be used. It may be
possible to improve the system considerably before the reactor is built
as a result of studies now in progress.

As now visualized, the reactor gases would be handled by one of two
methods (See Figure 12): 1) For product gas volumes below 500cc/min,
which would be the case for a major portion of the "in" time, the gas
would be diluted with 6 times as much air, to produce a non-explosive
mixture, and sent to a 3000 gallon underground holding tank for
radioactive decay. Ten days would be required for traversal of the
baffled interior of the holding tank, during which time the activity
would decay by a factor of 5 x 105 (Section 5 ). The gas emerging
from the holding tank (500cc/min, maximum) would be diluted with
10,000 cfm of air and blown up the building stack. With uniform mixing,
the gases emerging from the stack would have an activity of 2.5 x 10-4
microcuries.

2) For product gas volumes from 500cc/min to 2000cc/min (maximum
for 10KW operation), steam at 100°C is used for dilution, 4:1. The
resulting non-explosive mixture is passed through a stainless steel
wool-packed preheater, where a hydrogen-oxygen combining reaction is
initiated. The reacting gases are swept into a "converter" chamber
where the exothermic H-O reaction is controlled by cooling coils. The
total water vapor, both from the recombined H-O and the dilution steam,
is condensed and sent to a holding tank where the short-lived activity
decays. The small volume of non-condensed gases are then sent (a) to
the 3000 gallon holding tank for radioactive decay before atmospheric
dispersal, or (b) to a cooled activated carbon absorption trap where they
are absorbed until radioactive decay is adequate for atmospheric dispersal.


[page 47]

If at any time an unsafe quantity of radioactivity is found in the
gases being dispersed into the atmosphere, the reactor will be closed
down until the situation is rectified. If an inadvertent power flash
in the reactor should occur (a sustained high power is impossible) and
create suddenly a volume of product gases, these would be diluted with
air and swept to the holding tank where they would be held as necessary
for radioactive decay.


[page 48]

CONFIDENTIAL
III. REACTOR HAZARDS A. NORMAL HAZARDS.

In the routine operation of the reactor and its associated facilities,
certain hazards to personnel will exist. The situation is quite analogous to
that existing in an x-ray laboratory or in a chemicals manufacturing plant where
toxic gases, say, fluorine, is handled. In those and all similar situations,
safety or personnel is insured by (1) proper design of equipment, (2) adequate
monitors and safety devices and (3) continuous education and emphasis on safe
practices. The normal hazards of operating this establishment are listed below,
together with the means of insuring safety of personnel.

1. Radiation

Radiation may come from two sources: (1) open ports in the reactor
shielding from which a direct beam may emerge. Anyone entering the path
of such a beam would receive a dose of radiation of greater or lees magnitude,
depending on many factors. The beam of maximum possible intensity as it
emerged from a 3" hole at the surface of the shielding would contain

This beam would cause a radiation exposure over a 3" circular area of
approximately 2.0x10³R/second. At the wall of the reactor room, due to
attenuation and "inverse square" spreading of the bear, the radiation ex-
posure
would be about 1.3x10²R/second over a 12" circular area. If the beam

CONFIDENTIAL


[page 49]

strikes an object in its pathway, considerable amount of scattered
radiation over the reactor room may result. (2) Radioactive sources, e.g.
irradiated samples, etc., could cause irradiation of persons in the vicinity.

The following means, among others, will be followed as precautions
against excessive radiation exposure:


[page 50]

2. Radiochemical and Radiophysical Hazards.

In pre- and post-exposure handling of samples and specimens, considerable
manipulation of radioactive materials, largely beta and emitters, will be
involved. Handling, chemical processing, measuring, weighing -- all these
and similar operations may involve hazards of exposure, ingestion and personal
contamination. Various precautions will be followed to insure the safety of
personnel.


[page 51]

B. EMERGENCY HAZARDS.

In this category are listed inadvertent, unexpected, unplanned and ab-
normal
occurences and accidents which could or might result in personnel hazard
or area contamination of less than catastrophic proportions.

1. Leak or Rupture of the Reactor.

If a leak in the reactor occurs, radioactive liquid and radioactive
gases may be released. The reactor envelope is provided for just this
occurence. The released liquid and gas will be contained in the envelope.
The Liquid Salvage Line (Section II, D, 15) and a vacuum pump may be used to
withdraw the escaped liquid from the reactor envelope into shielded, "always
safe" containers. The gases may be pumped from the envelope via the
purge lines to absorption traps or to a gas holding tank and subsequent-
ly
to the stack for disposal.

When the liquids and gases have been removed from the reactor and
reactor envelope, the fluids and the contaminated parts of the system needing
repair will be handled as any other similar material. If the parts to be
repaired cannot be decontaminated; i.e., if the activity is due to the metal
itself, radioactive decay must be permitted or the part must be discarded.

The graphite blocks immediately below the reactor but included in the
envelope contain small drain holes to allow the fluid, in case of leakage,
to flow freely to the bottom of the reactor envelope. The possibility of a
sustained reaction in this region, however, was considered assuring that the
U235 was distributed uniformly throughout the graphite in the cylindrical
volume just below the reactor. Calculations show that such a reaction would
be impossible.


[page 52]

2. H2-02 Explosion.

The hydrogen and oxygen resulting from the radiation decomposition of
the fuel solution recombine with violence when ignited under certain con-
ditions
. Should this occur, pressures up to 20 atmospheres may result
if all the gas in a given volume engages instantaneously in the reaction.
The pressures actually resulting, which may be considerably less than 20
atmospheres, night cause rupture of the gas disposal tube or of the reactor
itself. Damage to the reactor envelope almost certainly would not occur.
Hence the most serious result would be a leak or rupture of the reactor,
which would be handled as above.

3. Inadvertent Removal of Control and Safety Rods.

The anticipated behavior of the reactor has been determined for the case
in which the control rods are removed completely from the reactor in a five
second period and also for the more extreme case in which the rods are re-
moved
instantaneously. In both cases it has been assumed that the reactor
fluid was at an initial temperature of 20°C.


[page 53]

Case I - Removal of Rods in Five Seconds

In this case it was assumed, due to the Blow removal of the rods, that
the delayed neutrons wore all of one period and group. The value used for
the percentage of delayed neutrons was 0.65% and for the period, 1/6 second.
This assumption results in a simple power equation which can be integrated
numerically to obtain the general reactor behavior for relatively rapid rod
removal. In those calculations it was further assumed that:

The above assumptions are in some cases unrealistic but all except (3)
and (5) are such that the most unfavorable conditions would be realized from
a safety standpoint.

The rods are assumed to be removed at such rate that the effective
multiplication constant increases linearly with time until the maximum
excess reactivity of 0.0157 is reached at the end of five seconds. The rods
are then left in this position indefinitely. (I.e., it is assumed that the
automatic rod releases do not operate.) The behavior of the reactor under
these conditions is summarized in Figure 13.

Numerical integration of the transient equations shows that the reactivity
increases with the sane rate as the rod removal until the "bubble" effect
becomes significant at about two seconds. At this point the power has


[page 54]

increased sufficiently to cause gas formation throughout the solution but
the temperature has increased too little to significantly effect the
reactor due to the large heat capacity of the system and to the short time
interval involved.

The power continues to rise until the rods have been nearly removed and
the bubbles are being formed at a rate sufficiently high to reduce k to unity.
This occurs at about five seconds. At this stage the temperature rise is such
that the temperature effect on k becomes important.

The reactor is now at the point where the sum of the temperature effect
and the "bubble" effect just equals the excess reactivity of 0.0157. The
temperature, however, continues to rise due to the inability of the cooling
system to dissipate all the heat generated. This increase in temperature
causes k to decrease, which lowers the power and the amount of gas evolved.
The temperature effect therefore gradually balances or cancels the excess
reactivity and the power drops to 10 kw and the temperature rises to approxi-
mately
95°C. This is below the boiling point of the sulfate solution and will
have no serious effect on the reactor components.

The power the reactor has reached at its peak is about 400 kw in this
exaggerated case. This is about forty times the maximum operation
level of the reactor and causes the radiation from the reactor to be high for
several seconds. The level of radiation at the surface of the reactor
vessel will be roughly:

The concrete shield enclosing the reactor is more than twice the thick-
ness
required for adequate protection at normal operation. The attenuating
ability of the concrete is therefore capable of affording adequate protection
from radiation intensities many times greater even than those released in the
emergency described above.


[page 55]

Case II. Instantaneous Removal of Rods

The case of instantaneous rod removal must be treated in a nuclear
different from that of Case I, due to the short reactor period resulting from
the excess reactivity of 0.0157. The "inhour equation" gives a period of 6 x 10-3 seconds for this reactivity.

The energy realized will then be given by:

where To is the period at 0.0157 excess reactivity, Po is the initial power
and t is the time. An increase in volume of 7.5% due to the "bubble
effect" will reduce the 0.0157 reactivity to zero. This takes in consideration
the effect of the reflector. Assuming that the gas forms bubbles immediately,
the power will rise for only 5.4 x 10-2 seconds.

The energy released during the power rise would be about 5.4 x 105
joules, or 1.3 x 105 calories. This means that the temperature increase
during the power rise time would amount to about 9°C. The temperature
would continue to increase, however, until the power dropped to 10KW (rate
at which heat is dissapated). The temperature would reach a maximum of about
95°C, which is well below the boiling point of the solution.

At the peak power of 105KW the flux densities at the reactor surface are
roughly 5.4 x 1016 [gamma]'s/cm²sec; 1016 fast neutrons/cm²sec. and 3 x 1016 slow
neutrons/cm²sec. The shield as designed is sufficient to protect personnel
in the vicinity during such burst of radiation.

It can be seen that at the worst there can be only a small rise in
temperature with the fluid possibly foaming suddenly into the space above
the liquid level and into the 1 1/2 exhaust tube. This could not result
in any damage to the equipment or personnel.


[page 56]

The results of the two cases are tabulated in Table 3.

TABLE 3
 Case ICase II
Energy Release8 x 105cal.8 x 105 cal.
Maximum Temperature95°C95°C
Pressure Increase<1 atmosphere <1 atmosphere
Maximum K1.0061.0157
Maximum Power Level400KW9 x 104 KW

4. Failure of Water Supply.

If a failure in the water supply should occur for any reason, the reactor
would suffer no ill effects. At any given setting of the control reds, the
temperature would increase and cause a decrease in the power level. This
would continue until the power level of the reactor became equal to the
dissipation rate of heat due to conduction through surfaces of the reactor.
This power level would have a value somewhat less than 1 KW. Thus, water
failure would result only in a reduction in the power level of the reactor.

On automatic control operation, however, the temperature would
increase to 95°C because, as the power level decreased, the automatic con-
trols
would move the control rods out in an effort to maintain the power
level constant. The power level would thus remain constant until a
temperature of 95°C was reached. Thereafter, regardless of the control
rod's position the power would drop to around 1 KW.

It should be made quite clear that the steady state level of operation
of the reactor, because of the negative temperature coefficient, is directly
dependent on the removal of heat from the reactor. With no heat removed


[page 57]

except that lost from the reactor surface, say, 1 KW, the steady level of
operation cannot be made to exceed 1 KW, regardless of the position of the
control rods. If 10 KW are removed by the cooling coils, then the reactor
will operate at 10 KW, etc.

5. Escape of Radioactive Gases from the Stack.

To evaluate the hazard resulting from release of radioactive gases
from the stack, it was assumed that the gases formed in the reactor in 10
minutes operation at 10 KW were released (over a 10 minute period) into the
stack stream from the 10,000 cfm blowers. This release normally could not
occur, since the reactor gases can reach the stack only through the 3000
gallon holding tank. A break in the gas conveying lines may be assumed to
cause release of this much gas into the ventilation stream to the stack, for
the purpose of this calculation, or an equivalent activity might be assumed
from some other source. The assumptions and calculations are shown below:

The potential radiation dosages resulting from 10 minute exposures
to such an assumed spherical air mass at various times after release are
listed in the table below. No distortion of the "cloud" by air movement
or dispersion of the radioactivity by diffusion is assumed.


[page 58]

Position Relative
to Cloud Chamber
Radiation Dosage Received in 10 Minute Exposures at
10 min.24 hours10 days, after release
Center250 R2. R0.2 R
Surface900.70.07
50 feet0.130.001Neg.
100 feet0.05Neg.Neg.

The effect of other releases of gaseous products of larger or smaller
volume activity may he compared with this calculation to obtain the maximum
possible effect. Inadvertent gas releases in the laboratories associated
with the reactor would probably occur from time to time, and these would be
swept through the ventilation system to the stack for dispersal into the
atmosphere. It is inconceivable, however, that gas volumes and activities
of the magnitude described above would ever be used in laboratory work,
hence any releases from this source would be on a many fold smaller scale.

A release of anywhere near this magnitude could occur only from the
reactor itself, by one of a series of circumstances such as the following:


[page 59]

6. Effect of Water Removal for the Cooling Coils.

The reactivity of the reactor increases if water is removed from the
cooling coils. This is due to the fact that, in a well moderated system
such as the reactor, the removal of the water does not materially change
the moderation but does remove the neutron absorption of the amount of
water (H) involved. Calculating on the basis of actual experiments per-
formed
at Los Alamos, and the cross-sections of the elements in the water
and the stainless steel of the coil, a value of 0.3 grams of U235 is found
to be the effective equivalent of each "foot" of the water. The total
removal of the water from the cooling coils would be equivalent to
adding 9 grams of U235 to the reactor. The control rods could easily
handle this increased reactivity.

7. Addition of Water or U235 Solution to the Free Volumes in the Reactor.

The fuel solution is at optimum moderation for minimum critical mass
in the dimensions of the reactor. If appreciable quantities of water are
added or removed, to make the solution more or less diluted, the reactivity
becomes less, or more U235 is required to produce criticality. Thus, filling
the free volumes of the reactor, i.e. above the liquid surface, Th the con-
trol
rod scabbards, etc., would reduce the activity of the reactor, as ex-
plained
in Section 6 above.

Filling these volumes with additional uranyl sulfate solution would
increase the reactivity of the reactor. About 1500 cc of solution containing


[page 60]

at optimum concentration, 100 grams of U235, could be placed in the reactor
above the normal liquid level. Eight more grams of U235 could be placed in
the re-entrant "glory hole", and thirty grams inside the cooling coils (if
the water were first removed). Thus, total U235 of about 138 grams could
be placed inside the reactor's free volumes if one were deliberately intent
on creating the greatest possible excess reactivity. The two reactor rods
together would more than equal this increased uranium content, but one
alone could not prevent activity. The reaction resulting from additions of
this kind would be temperature-limited, as described in Section 3 above.

C. CATASTROPHIC HAZARDS

In this category are included those unforseen and unplanned events and
accidents of such violent proportions that the reactor and building may be wrecked
and the lives of persons in the vicinity are endangered. Only two means can be
visualized by which a catastrophe could occur: (1) earthquake or other act of
God, and (2) sabotage by (non-nuclear) explosion.

In this category would also be listed any violent and uncontrolled
nuclear reaction resulting in explosion which could inadvertently occur.
None can.

If catastrophe should occur, there is only one hazard, which could ensue,
beyond the event which causes the catastrophe, namely contamination of the area
with radioactive liquid and gases.

If the dispersal is into the reactor room only, with the walls of the
room remaining intact, the liquid would spill or be washed into the drains.
These, however, do not go to the city sewer system except via holding tanks in
which radioactive materials are retained until safe for disposal (Section II, C).


[page 61]

Hence, the city sewers would not become contaminated, and perhaps some or much of
the fuel could later he salvaged. The gases will be retained in the reactor room,
or will escape slowly, or may he dispersed via the blowers and stack to the at-
mosphere
. Clean up of the dispersed liquid would constitute a terrific problem.
The gases released would probably not exceed in activity those described in the
preceding section on emergency hazards, and their disposal would be handled in
similar.

If the reactor room as well as the reactor is demolished, the fuel solu-
tion
is likely to be thrown out upon the ground somewhere in the vicinity, where
it will soak into the ground and be lost. The area immediately adjacent to the
site of the liquid spill, and all objects on which the liquid was thrown would
be highly contaminated. Subsequent hazard to personnel due to the liquid, other
than that occurring during the catastrophic event, would be slight. Due to the
topography and soil and subsoil formations, the chances of the spilled liquid
entering an underground flow channel and reaching eventually the city water
supply are vanishingly small.

In a catastrophic event which demolishes both the reactor and the
reactor room, the relatively small volume of gases and vapors, oven though a portion
of the solution were vaporized in the explosion, would he violently and widely
dispersed upward and outward over the area. The resulting "cloud" would drift
away from the site with the wind. The size of the cloud and its intermixing
with air would depend on the violence of the explosion. If the explosion destroys
the massive reactor shielding and simultaneously demolishes the walls of the
building, the radioactive gases would he blasted into a large, highly turbulent
volume. no less certainly than 1 or 2 hundred thousand cubic feet. The reactivity
in this cloud would result from vaporization and dispersal of a portion of the re-
actor
solution, perhaps one tenth of the total volume. This amount would contain
an initial activity of about 1.8 x 10³ curies.


[page 62]

Thus the radioactive cloud from a catastrophe of this sort would result in a
"spherical cloud" having the same size and 50 times the activity of those des-
cribed
in Section III, B, 5. In this case, the cloud would begin its migration and
dispersal at ground level rather than at 100 feet or so in the air.

A considerable number of precautions against sabotage have been in-
corporated
into the reactor design:


[ Figure 1]

FIGURE 1. TOPOGRAPHIC MAP OF RALEIGH


[ Figure 2]

FIGURE 2. NORTH CAROLINA STATE COLLEGE CAMPUS


[ Figure 3]

FIGURE 3. TENTATIVE SKETCH PHYSICS RADIATION LABORATORY


[ Figure 4]

FIGURE 4. REACTOR SHIELDING SKELETON


[ Figure 5]

FIGURE 5. COMPLETE REACTOR SHIELDING ASSEMBLY


[ Figure 6]

FIGURE 6. STAINLESS STEEL REACTOR, ONE HALF SIZE


[ Figure 7]

FIGURE 7. REACTOR ENVELOPE AND SAMPLING TUBE


[ Figure 8]

FIGURE 8. HORIZONTAL CROSS SECTION THROUGH THERMAL COLUMN


[ Figure 9]

FIGURE 9. EXPOSURE PORTS


[ Figure 10]

FIGURE 10. LIQUID LEVEL INDICATOR


[ Figure 11]

FIGURE 11. DECAY OF GASES PRODUCED DURING ONE MINUTE OF OPERATION


[ Figure 12]

FIGURE 12. GAS DISPOSAL SYSTEM, SCHEMATIC DIAGRAM


[ Figure 13]

FIGURE 13. FIVE SECOND ROD REMOVAL EMERGENCY


Notes:

a1"A Curriculum in Nuclear Engineering", by Clifford Beck, March 25, 1950

a2"Proposal of a Nuclear Reactor at North Carolina State College", by Clifford
Beck
. July 1949, revised March 1950.

a3Contract #AT-(40-1)1032, June 5, 1950.

b1Dr. Ralph Fadum, Head of Civil Engineering Department, in private correspondence

c1Fig. 11 relates to operation at 10 KW.