Further Design Features of the Nuclear Reactor at North Carolina State College



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FURTHER DESIGN FEATURES
OF THE
NUCLEAR REACTOR
AT
NORTH CAROLINA STATE COLLEGE

Clifford Beck
A. C. Menius, Jr.
R. L. Murray
Newton Underwood
A. W. Waltner
George Webb

PHYSICS DEPARTMENT
SCHOOL OF ENGINEERING NORTH CAROLINA STATE COLLEGE
of the
UNIVERSITY OF NORTH CAROLINA
Raleigh North Carolina

January, 1952



The material presented in this report has
resulted from the combined efforts of various
members of the Physics Department. Almost
all members of the department, in one way or
another, have made noteworthy contributions
to the reactor project. Members of the other
departments in the School of Engineering,
have also made significant contributions. The
assistance of these men and the value of their
contributions are gratefully acknowledged.

The Authors



TABLE OF CONTENTS

LIST OF FIGURES AND TABLES



FURTHER DESIGN FEATURES
OF THE
RALEIGH RESEARCH REACTOR
AT
NORTH CAROLINA STATE COLLEGE

INTRODUCTION

A proposal to construct and operate a small nuclear reactor on the campus
of the North Carolina State College, at Raleigh, N. C., and the general design of
a reactor which would be suitable for this purpose have been presented in previous
reports.a1a2a3 The Atomic Energy Commission has expressed approval of this
projecta4 and of the general design of the proposed reactor. It was recognized at
the time of approval, of course, that a great deal of work yet remained to be done
on the details of the design of the reactor before construction could begin.

Additional general plans of the reactor facility and further details of the re-
actor
design and of the auxiliary systems have now been completed. In accomplish-
ing
this, the staff at N. C. State College have drawn heavily on the experience and
ideas of the Los Alamos and Oak Ridge staffs. Designs of numerous component
systems have been borrowed from the successful Los Alamos water boiler and
from the reports on reactor studies at Oak Ridge, and have then been adapted to
particular conditions and needs of the Raleigh Research Reactor. It has not been
the intention of the State College staff to produce a novel design of a reactor or to
make original contributions to reactor technology. Rather, it has been the purpose
to design and construct as simply and quickly as possible, a safe, flexible nuclear
reactor with maximum adaptability to instructional and research purposes. The
staff of N. C. State College is therefore heavily indebted to the A. E. C. and to the
technical personnel of Los Alamos and Oak Ridge for the many suggestions and
ideas which they have contributed to the Raleigh Research Reactor project with
unfailing generosity and cooperation.

The present report presents a general description of the Reactor Building,
plans of the Reactor components, further discussion of potential hazards which
may be involved, and the anticipated start-up procedures and operating policies.


[page 2]

I. THE. REACTOR BUILDING A. General.

The Raleigh Research Reactor is to be located in the southern half of the
Court of Ceres, an open quadrangle near the center of the North Carolina State
College
campus. The Court of Ceres lies between the buildings of the Engineer-
ing
School and those of the Agricultural School. The location chosen for the re-
actor
, Site B as described in the report "Program Administration and Installa-
tion
Design of the Nuclear Reactor Project"a2 (p. 11 and Fig. 2), is one of the
highest locations on the college campus and in the entire city of Raleigh. The
buildings housing Physics, Chemistry, and Engineering Research, whose activi-
ties
particularly relate to the functioning of the reactor, are immediately adja-
cent
to the quadrangle in which the reactor is being constructed. Thus, for
several reasons the site chosen seems to be an advantageous one.

The plan of the Reactor Building is shown in Figures 1 and 2. A sketch of
the external appearance is shown in Figure 3. The octagonally shaped reactor
assembly is below ground level in the center of the 57 foot diameter Reactor
Room. This room, in turn, is in the center of the building. On three sides of
the room at ground level, are fourteen laboratory rooms to be used for instruc-
tion
and research. Underneath these laboratories, at the floor level of the
Reactor Room, are five additional utility or laboratory rooms.

On the fourth side of the Reactor Room are the Control Room and the Ob-
servation
Room, both being separated therefrom by masonary walls 12" thick
and water-filled glass windows, 8" thick. In front of the Observation and Con-
trol
Rooms is an entrance lobby and four offices.

The photographic dark room, the counting laboratory and the Control
Room are air conditioned. The entrance lobby and the offices are provided
only with ordinary window ventilation. A ceiling exhaust fan furnishes ventila-
tion
for the Observation Room. Except for the Control Room, these areas will
be generally accessible to visitors and observers.

It is not anticipated that radioactive contamination of these areas of the
building will occur at any time. Plumbing facilities here are connected directly
to the city system in the usual manner. The remaining areas of the building,
i.e., the laboratories on upper and lower levels and the Reactor Room, may
become contaminated with radioactive materials and therefore special ventila-
tion
and sewer facilities are provided.


[page 3]

B. Ventilation and Heating System for the Laboratories.

Separate systems are provided for supplying filtered air to and removing air
from the laboratories. The intake air system is made up of two parts: one for
the west side and one for the east side of the building. In each of the two parts,
air is drawn through a bank of disposable paper filters located under the overhang-
ing
eaves of the roof, near the front of the building. Intake blowers are mounted
above the ceiling of the locker rooms and from these the air is distributed by ducts
to each laboratory. The blowers in the supply system can be run either at full or
at half capacity. In normal operation, each blower will operate at half capacity,
which is 6,275 cfm. At full capacity, each system can supply 12,500 cfm, or a
total of 25,000 cfm.

The air is delivered to each laboratory through a control damper mounted
on the inside wall of the room. In normal operation, the air is changed once every
10 minutes.

Providing comfortable temperature conditions in cold weather, with air flow-
ing
into and out the building at the above rate, would tax the capacity of usual build-
ing
heating systems. For this building, therefore, a dual system is used Heat
exchangers in the main air supply duets insure a minimum air temperature of 65oF.
In addition, coils of hotwater pipes mounted in the ceiling or walls of each room
provide radiant heat to the extent needed for comfort.

The Exhaust Ventilation System withdraws the air from each laboratory
through the hood in that laboratory and through the filter in the exhaust line under
each hood, then through a system of ducts underneath the floor, to the filter room
at the rear of the building. Air is supplied to and removed from the Reactor Room
through louvered ventilators mounted in the wall of the room. Ducts from the ex-
haust
ventilators of this room connect to the remainder of the Exhaust Ventilation
System. Two 12,500 cfm centrifugal blowers, arranged in parallel, draw the air
through the bank of filters and discharge it through a 48 inch diameter, 110 foot
stack located at the rear of the building. In normal operation, only one of the two
blowers is used. If desired, however, both blowers can be used simultaneously,
in which case, 25,000 cfm are withdrawn through the exhaust system. At times
it may be necessary to maintain the flow of air up the stack, when ventilation of
the building is not desired. To permit this, a "by-pass" from the outside wall at
the rear of the building to the blower intake is provided. If this by-pass is com-
pletely
or partially open the full demand of the blowers can be satisfied with no
air, or only part of the total, coming from the building itself. Operation of the
by-pass can be either automatic or manual.

The hoods in each room are of a special, downdraft design, having removable
filters located in the exhaust duct under each hood.


[page 4]

C. Sewerage System for the Laboratories.

It is not intended that any quantity of radioactivity above maximum concen-
tration
permitted by A.E.C. regulations shall escape into the drain lines of the
building. Short-lived radioactive materials above this concentration will be stored
until sufficiently decayed, and long-lived materials above this concentration will be
stored for disposal in other ways. Due to emergencies, accidental spillage or
misoperation, however, above tolerance concentrations of activity may occasionally
escape into the drain lines. For this reason, therefore, no drain lines from the
work areas of the building connect to the city sewer system, except through a radia-
tion
monitored tank system.

All drain lines from the Reactor Room and the work rooms on the lower level
discharge into a sump near the air filter room. A pump lifts the material from the
sump to the drainage system of the laboratories above. The liquid at the sump is
continuously monitored, and if excessive amounts of activity are found, the material
is not pumped to the drainage system above. It should be necessary to use the drains
on the lower level only seldomly and hence the sump pump will need to be used only at
infrequent intervals.

The drainage system from the ground level laboratories passes through
system of monitors, holding tanks, and valves designed to prevent discharge of
radioactive materials above concentrations permitted by A.E.C. regulations to the
city sewer system. (Fig. 4)

In normal operation, laboratory drainage passes a radiation monitor and a
cut-off valve into one of two 550 gallon tanks, arranged in parallel. The effluent
line of this tank is halfway from top to bottom, so that the tank is always half full.
As liquid leaves the tank it passes a second radiation monitor, located in the exit
line, and then flows through an automatic, pneumatically operated valve to the city
sewer system. When desired, the exit line from the midpoint of either tank can be
closed and the tank becomes a holding tank with 275 additional gallons of capacity.

In normal operation, the activity of materials discharged down the drains
will be positively controlled within permissible limits before release. The radia-
tion
monitors and the hold-tanks serve as secondary defenses against unsafe
amounts of radioactivity reaching the city sewer system in case activities of above
tolerance levels are inadvertently released into the drains.

The monitored holding tank system described above is intended to operate
as nearly as possible as follows: If an amount of activity above the maximum
permissible level goes down the drain, it passes the first radiation monitor, which
sends a signal to the Control Room and also sends a signal to the pneumatically
operated valve in the exit line of the hold-tank. The valve closes and the radio-
active
material is caught in the tank. The operator then opens the valve to the
second tank and this one is used in the drainage system while the decision is made
as to the method of handling the material caught in the first tank. The radiation
monitor in the exit line of the flow tank serves as a double check on the first moni-
tor
. It likewise acts to close the exit valve of the hold-tank if activity of excessive


[page 5]

level reaches that point.

In actual practice, it may prove quite difficult to obtain radiation monitors
capable of performing with reliability as described above, because the levels of
activity intended to be discharged are extremely low. (Actually, the exact values
of the maximum permissible concentrations have not yet been determined. Fur-
ther
conferences with the A.E.C. on this matter will be necessary.) Therefore,
in addition to the plan outlined above, in which the best monitoring procedures
available will be used, spot checking of liquid samples withdrawn from various
points in the system and other means as necessary will be instituted to insure
that activities above safe levels are not discharged. If all possible methods of
continuous operation of the system proves unsatisfactory, the two tanks can be
used entirely as hold-tanks. That is, material can be discharged into one tank
until it is full, and then the flow be diverted to the other. Meanwhile, sampling,
monitoring and analysis of the material in the first tank should be made to de-
termine
whether the material should be discharged to the sewer, held for radio
active
decay, or pumped to storage tanks, etc.

Each tank is provided with a gas vent line connected to the exhaust system
of the building, a valved overflow line connecting to the city sewer system, a
valved drain-line (normally closed) connecting to the city sewer system for com-
pletely
draining the tank, and a vertical 4" pipe through which sampling or inside
monitoring of the tank contents can be achieved.


[page 6]

II. REACTOR DESIGN A. General.

Since 1944, the Los Alamos Scientific Laboratory has operated a low-power
nuclear reactor of the "water-boiler" type.b4 At first the reactor was operated
at a level below 1 kilowatt, then in step-wise changes in the design over a period
of 6 years, at 5, 10, and now 25 kilowatts. Operation at each level has yielded
knowledge and experience on the stability, behavior and use of a reactor of this
type. With added experience, increased confidence and satisfaction with the basic
principles on which the unit operates and its inherent safety, adaptability and
utility have been gained.a5

The reactor being built by N. C. State College is also of the homogeneous,
water-boiler type, and its fundamental principles of operation are identical with
those of the Los Alamos unit. Such differences in details of design have been in-
corporated
in the Raleigh Research Reactor as would be expected to adapt it most
satisfactorily to its proposed location on the college campus, and its proposed
usage in research and instructional programs.

The Raleigh Research Reactor will contain on the order of 800 grams of
U235, of 90% isotopic enrichment. The uranium will be in the form of uranyl
sulfate in water solution, and will be contained in a stainless steel cylinder
10-3/4 inches (i. d.) in diameter and 11 inches high (inside dimensions). The
cylinder will be enclosed in 20 inches of purified graphite reflector and this, in
turn, will be surrounded by six feet of a special, high density concrete. (Fig. 5)
The graphite reflector is extended out in one horizontal direction to form a
thermal column five feet square and five feet long. Four internal cooling coils
in the fuel cylinder provide heat removing capacity to enable the unit to operate
at a power level of 10 kilowatts. Besides the cooling system, instrumentation
for operation and control and a "gas disposal system" for handling the gaseous
by-products from the reactor, are the chief auxiliaries.

B. Concrete Shielding.

The concrete shielding performs two essential functions: Absorption of
radiation from the reactor and protection of the fissionable fuel from danger of
theft or sabotage. The shielding must be arranged so that samples may be placed
inside for irradiation, or radiation beams may be allowed to emerge for external


[page 7]

use. Also, the shielding must permit convenient access to the internal components
for repair and maintenance.

For absorption of radiation, the concrete is made 6 feet thick, and is composed
of special ingredients.a6 To increase the density above that of ordinary concrete and
hence enhance its absorption of gamma rays, Barium Sulfate (Barytes Ore) is used
as the coarse aggregate. A boron containing ingredient (finely ground colemanite
ore) is used as the fine aggregate. The boron increases neutron absorption in the
concrete and hence neutron activation of the shield is reduced. The boron content
of the finished concrete is 1% by weight.

Final composition of the concrete, per cubic yard, is:

4200 lbs. Barytes - 95% between 1/16" and 3/4" size
5% smaller than 1/16"
423 lbs. Colemanite ore - 100% larger than 100 mesh sieve
size and smaller than 20 mesh
sieve size.
882 lbs. Portlant cement - Type 3.
51.7 gallons water.

The colemanite is soluble in water, hence considerable difficulty is encountered
in its use. There appears to be a competition for the water between the dissolving
action of the colemanite and the normal process of cement setting; the setting of the
cement is delayed enormously. The barytes appears also to be sufficiently soluble
to have some effect on the setting process. Two proceduresa7 were found to be help-
ful
: (1) Using an absolute minimum of water, and (2) adding the colemanite in a
second mixing process after some of the primary setting processes had commenced
(40 to 45 minutes later). The concrete made in this way was found to have almost
double the strength of normal concrete. An overall density of 3.4 g/cc was achieved
in contrast to 2.4 g/cc for that of ordinary concrete.

When assembled, the concrete shield is octagonally shaped in horizontal cross-
section
, 17 feet across. A rectangular cavity in the center is provided for housing
the reactor, the reflector and the auxiliary components. A 36" concrete slab covers
the top of the assembly. (Figs. 6, 7) The shield can be partially disassembled by
removal of various interlocking concrete blocks making up the assembly. The con-
crete
blocks range in weight up to about 6 tons. To disassemble the shield, the


[page 8]

"cap-stones" on top of the assembly must be removed first, and then the inter-
locking
blocks underneath can be lifted out one by one. (Fig. 8)

When all the portable concrete blocks are removed, there remain the two
massive sides of the octagonal shield, separated by a five-foot gap. Thus, access-
ibility
is provided for the initial assembly of components inside the shield and for
subsequent maintenance and repair. (Fig. 9)

There are 11 exposure ports extending through the shielding and reflector
to the surface of the reactor or across the thermal column. All external open-
ings
are so arranged that radiation leakage is prevented, and unauthorized access
to the interior of the shield is prohibited. Each port (Fig. 10) is closed externally
by a combination-lock safe door, the frame of which is imbedded in the shielding
concrete. The outer 18" of the steel lined port is 7 inches in diameter. From
this point to the inner end of the port the diameter is 6-1/2 inches. This one
quarter inch annular offset in the port prevents escape of radiation through the
crevices surrounding the concrete plugs which fill the port when it is not in use.

One special port consists of a 1 inch vertical tube extending from the top
center of the concrete shield, downward to the reactor and re-entrant into the
center of the reactor. Small samples may thus be placed at the center of the re-
actor
for exposure to the highest possible flux.

Attenuation of radiation by the shielding, and the levels of radiation ex-
pected
on the outside surface of the shield during reactor operation, are shown
in Table 5, in Section III, below.

C. The Fissionable Fuel.

The reactor at Los Alamos has, at different times, operated with water
solutions of uranyl sulfate and of uranyl nitrate as the nuclear fuel. From the
standpoint of neutron absorption, the sulfate has some advantage. From the
standpoint of solubility and corrosion rate, the nitrate appears to have a slight
advantage, particularly at high temperatures and pressures. It appears, how-
ever
, that either material would serve satisfactorily as the fuel for the proposed
reactor, and there is at present no strong reason for choosing one instead of the
other.

Uncertainty in the exact chemical composition and geometry of the fuel con-
tainer
, and of the surrounding materials, the presence of a number of attached
and re-entrant tubes, and other factors of uncertainty make it impossible to cal-
culate
the exact amount of U235 needed in the reactor. Calculating from the
known critical mass of the Los Alamos reactor, making allowance for the differ-
ence
in geometry (sphere to cylinder), difference in size and arrangement of
cooling coils and control rods, and estimating the effect of all other different
features, it appears that 715 grams of U235 in uranium of 90% isotopic purity,
as uranyl sulfate in water solution, are required to produce criticality at room


[page 9]

temperature. To this must be added the amount required to overcome the negative
temperature effect at operating temperature and the "working excess" needed in
experimental procedures.

A tabulation of various calculated quantities and characteristics relating to
the nuclear fuel is presented in Table 1.

TABLE 1 - CHARACTERISTICS OF THE REACTOR FUEL

Estimated critical mass at room temperature715 gm U235
Added U235 for temperature coefficient55 gm
"Experimental Excess"20 gm
Total U235 content790 gm
Solution density (9% UO2SO4 by weight)1.08 gm/cm3
Hydrogen to U235 atom ratio450

D. The Fuel Container and Reactor Envelope.

Type 347 stainless steel is used throughout the reactor system wherever con-
tact
with the fuel or its vapor is a possibility. Welding is used whenever possible
in preference to other means of connection.

A volume of 14 liters is provided in the reactor cylinder for the fuel solution.
In addition to the fuel volume, the reactor cylinder (Fig. 11) contains numerous
connecting and re-entrant tubes and an empty space at the top to allow for frothing
and expansion of the fuel solution.

Data relating to the fuel cylinder are contained in Table 2 below:

TABLE 2 - DATA RELATING TO FUEL CYLINDER

(347 stainless steel)
Diameter o.d.10-7/8 inches
i. d.10-3/4 "
Height, outside11-1/8 "
inside11 "
Wall thickness1/16 "
Weight of steel in walls2.3 kg.
Total inside volume15.6 liters
Inside volume occupied by re-entrant tubes0.71 liters
Weight of steel in inside re-entrant tubes1.9 Kg.
Liquid depth9.92 inches
Liquid volume14.0 liters
Depth of space above liquid1.08 inches
Volume of space above liquid0.90 liter


[page 10]

In case a leak should occur in the fuel cylinder, radioactive vapor or liquid,
or perhaps both, depending on the location of the leak, would escape into the sur-
rounding
space. To minimize exposure hazards resulting from an event of this kind,
the reactor is enclosed in an envelope of 1/16" aluminum. (Fig. 12)

All tubes, control rod sheaths, and thermocouple leads connect to the reactor
on its top surface, and project vertically upward inside the reactor envelope. The
motor drives and electromagnetic couplings of the control rods are located inside
the top portion of the reactor envelope. Tubes and electrical wires leave the en-
velope
through seals in the walls.

E. Reflector, Thermal Column, Lead Shielding.

The reactor envelope which is snugly fitted to the reactor is, in turn, en-
closed
in 20 inches of high purity graphite which serves as a neutron reflector.
Calculations show that a 20 inch layer of graphite around the reactor core is 90% as
effective in reducing critical mass as an infinitely thick layer. Two to four inches
of lead are placed around the graphite reflector, inside the concrete shielding, as
a primary gamma ray shield. (See Section III for calculations of attenuation.)

On one side of the reactor, outside the lead shielding, an additional 5 foot
cube of graphite forms the thermal column. (Calculated values of thermal neu-
tron
fluxes are listed in Section III, C.) Several exposure ports penetrate into
and across the thermal column to provide means of using the neutrons for ex-
perimental
purposes.

A layer of lead four inches thick is placed across the end of the graphite
column. This is followed by a layer of boron to absorb the neutrons reaching
that point. The boron, in the form of finely. ground boron carbide, is impregnated
into a layer of paraffin 3/4" thick, in sufficient concentration to form a boron
layer of 3 gms/cm2 across the end of the column. A layer of concrete 12 inches
thick is placed outside the boron layer.

Specially prepared graphite, obtained from the A.E.C. at Oak Ridge, is used
in the reflector and thermal column. Data relative to the graphite are presented
in Table 3, below.

TABLE 3 - DATA ON GRAPHITE USED IN REFLECTOR AND THERMAL
COLUMN


[page 11]

The graphite portions of the assembly are built up of successive layers of
graphite bars 4 inches square in cross-section and in lengths up to 48". All joints
are fitted to ± 0.002".

F. Cooling System.

Due to the negative temperature coefficient of nuclear reactivity of the re-
actor
system, the steady state power level at which the reactor will operate is
limited by the rate of heat removal from the solution. If less heat is being re-
moved
than is being released, the temperature rises and, as a result of the ther-
mal
expansion of the fuel and the creation of gas or vapor bubbles, the reactivity
is reduced so that a heat balance is established.

Dissipation of heat from the reactor through the external walls of the vessel
will hardly exceed half a kilowatt. For operation at higher levels, therefore,
auxiliary cooling of some sort must be provided. The system devised consists of
four symmetrically arranged coils of 1/4" (i. d.) stainless steel tubing immersed
in the reactor fuel. One gallon per minute of refrigerated water flows through
each coil. A 7 foot length of each coil is immersed in the solution. The wall
thickness is 1/32 inch.

Water from the city mains flows through refrigerated coils in the air con-
ditioned
system, through control valves in the Control Room, to a distribution
manifold inside the reactor assembly and thence to the four reactor cooling coils
(Fig. 13). The coils are adjusted for equal flow before installation, but after in-
stallation
, flows in the individual coils are not measured or separately controlled.
Temperatures are measured and recorded at the inlet manifold and at the exit of
each coil.

In its passage through the reactor, the cooling water becomes somewhat radio-
active
. Calculations of the induced activity to be expected indicate that the total
amount will not be large and most of this will be short-lived. (Table 4) The
normal time required for the transit of an average water molecule through the re-
actor
is 1.1 seconds, and the average thermal neutrol flux is about 4 x 1011. For
calculation of the data presented in Table 4, a transit time of 2 seconds and a flux
of 1012 thermal neutrons were conservatively assumed. Average analyses of
water in the Raleigh system were obtained from the City Public Utilities Depart-
ment
.


[page 12]

TABLE 4 - ACTIVITY INDUCED IN THE REACTOR COOLING WATER

Target
Isotope
 Form of
the Im-
purity
 Avg.
Conc.
ppm
 Active
Isotope
 Half
Life
 [lambda]
sec-1
 [sigma]a
barns
 Abund. of
Target
Isotope
(%)
 N
cm-3
 Result.
Activity
d/sec/cm3
Si30 SiO2 9.8 S31 2.7h 7.1 x 10-5 0.12 3.1 1.0 x 1016 0.017
Mg26 Mg 1.4 Mg27 9.58m 1.2 x 10-3 0.05 11.3 5.3 x 1015 0.64
Na23 Na+K 4.2 Na24 14.9h 1.3 x 10-5 0.6 100 1.4 x 1017 2.2
S36 SO4 11.0 S37 5.0m 2.3 x 10-3 0.14 0.0136 5.0 x 1013 0.032
Cl35 Cl 4.9 S35 87.d 9.2 x 10-8 0.34 75.4 1.2 x 1017 10-2
Cl37 Cl 4.9 Cl38 38.5m 3.0 x 10-4 0.6 24.6 4.0 x 1016 14.
F19 F 0.1 F20 12s 5.8 x 10-2 0.009 100 3.35 x 1015 3.5
O18 H2O 106 O19 29.4s 2.4 x 10-2 0.0002 0.204 6.8 x 1019 653

Negligible activities were found for a number of additional elements likely to be
present. These are:

Fe54, Fe58, Ca44, C13, S34, Cl35, N15


[page 13]

From this table, it is seen that the O19 activity (0.018 micro curies per ml),
initially, is dominant. Since the half life is only 29.4 seconds, however, this
activity will be reduced below 1 disintegration per ml/sec in 5 minutes, and thus
no hazard would be anticipated.

The next most significant activity is that of Cl38 (3.8 x 10-4 micro curies/ml)
which is of the order of permissable concentration for drinking water, and hence no
hazard would be expected to result from discharge of this material into the sewer.

Inasmuch as the power at which the reactor operates is intimately associated
with the rate of cooling, an interlock with the control rod system is provided
which insures that the cooling water is flowing before the reactor can be operated.

The exit line carrying the cooling water from the reactor passes through an
underground, concrete shielded trench under the building to an underground tank
of 250 gallon capacity located outside the building, and from there to the city sewer
system. Radiation detectors continuously monitor the activity of the water as it
enters and leaves the tank. The same plan of operation will be followed for this
system as that described in Section I, C, for the laboratory drainage system.
As in the drainage system, an attempt will be made to use a continuous flow method
of discharging the reactor cooling water, through a monitored tank, to the city sewer
system. But if the level of radiation proves too high, or if reliable monitors cannot
be found for the low level, of radiation involved, then another tank will be installed
and the two will be used alternately as hold tanks The cooling water will then be
discharged batchwise to the city system after sampling and analyses of each batch.

G. Gas Disposal System. 1. Hydrogen-Oxygen Recombination.

When the reactor is operating at 10 Kw, from 1500 to 2000 cc of gas per
minute are evolved. More than 99% of this gas volume will consist of gases result-
ing
from radiation decomposition of the fuel solution. A very small fraction of the
gas volume will be fission product gases which, though negligible in volume,
greatly complicate the disposal problem, because of their high radioactivity.

Various possible methods of disposing of the gases were briefly outlined in
the previous report, "Program Administration and Installation Design of the Nuclear
Reactor Project at N. C. State College." Since that time a great deal of study, at
Oak Ridge and Los Alamos, as well as N. C. State College, has been devoted to
this problem.

The scheme developed by Los Alamos and used thus far with complete satis-
faction
and success (over an operating period of 20,000 kw-hr) has been adopted
for the Raleigh Research Reactor. Modification in details of arrangements and
components has been incorporated for adaptation to the new reactor but the funda-
mental
principles of operation have remained unchanged. The Los Alamos system


[page 14]

is based on a very ingenious combination of ideas and operating principles which
achieves freedom from explosion hazard, simplicity of apparatus and mechanical
components, low maintenance requirements, and satisfactory handling of radio-
active
gases.

The gas disposal apparatus as adapted for the Raleigh Research Reactorc1
consists of a closed circulating gas system, with a small exhaust gas bleed to the
stack and a small inlet make-up flow. The flow path (Fig. 14) includes, in se-
quence
, the empty space in the reactor above the fuel surface, a condenser cooled
with refrigerated water, a steel-wool-filled filter, a circulating blower, a
platinized-alumina catalyst bed, a water cooled steam condenser, and a pipe con-
nection
back to the top of the reactor. The system has a volume of 8000 cc, which
is filled with air (94%), and gases from the reactor (6%). Fifty to one hundred ml
per minute of mixed gases are bled from the circulating system through a series
of hold-up tanks to the stack.

A small volume of non-condensable gases from the reactor (SO, SO2...)
plus a small inflow of make-up air to the system is sufficient to maintain constant
circulating inventory. The inbleed of make-up air occurs to some extent around
the pump shaft, and the additional amount required is admitted through a small
adjustable leak located on the inlet side of the pump.

The essential features of this system are:

The details of construction and arrangement of the chief components in the
gas disposal system and other details are shown in Figures 15, 16. and 17.


[page 15]

2. Disposal to the Atmosphere.

About 50 to 100 ml per minute of gas are bled from the gas circulating-
recombining system during periods when the reactor is in operation. The point
of withdrawal (Fig 14) is located immediately downstream from the exit end of the
steam condenser where the H2-02 content is lowest.

The gas bled from the system cannot be discharged directly into the atmos-
phere
, because of its radioactivity. A hold-up system must be provided to delay
discharge into the atmosphere sufficiently to permit radioactive decay to a safe
level. Calculations showing amounts of gaseous fission products expected, their
decay characteristics, and the adequacy of the proposed hold-up system are pre-
sented
in Section III, C. Description of the system itself follows.

The plan of the gas withdrawal system is shown in Fig. 18. A 1/4" stain-
less
steel tube leads from the bleed point of the circulating system downwards
within the reactor shielding, then horizontally under the floor of the reactor room
and out underneath the building to a series of gas-holding tanks immersed in an
underground water tank (for shielding) outside the building The bleed gas is
drawn through the holding tanks by a small pump, and is discharged from the
pump into the 12,500 cfm ventilation exhaust stream from the Reactor Building.

A constant flow of gas is maintained through the system by a suitably sized
critical flow orifice located between the holding tanks and the exhaust pump. A
water trap, through which the gas must bubble, is located between each holding
tank. Eight-hold-up tanks in series, each having 100 gallon capacity, are, pro-
vided
. It is estimated in Section III, C, that the attenuation achieved in a series
of eight such tanks is on the order of 700 for Xenon133, which is the most
troublesome gaseous fission product. By-pass valves are arranged so that any
tank can be removed from service without disrupting operation of the others.

H. Instrumentation and Control. 1. General.

The instrumentation system of the reactor facility has four functions:
(1) to indicate and record the level of radiation flux and the rate of increase or
decrease of the neutron flux in the reactor; (2) to provide data for and means of
safe manual and automatic control of the reactor; (3) to provide safety mechan-
isms
which insure reactor operation within predetermined limits; (4) to provide
area and facilities monitoring to safeguard personnel from radioactivity
hazards and to prevent inadvertent release of radioactive materials.

The first three of these functions are performed by the instrumentation
system provided for the reactor itself. In general plan, this reactor instrumen-
tation
system consists of fission - or ionization-chamber sensing elements
located in the graphite reflector of the reactor assembly, which are connected


[page 16]

by coaxial cables and power lines to their respective power supplies, amplifiers,
indicating meters, recorders, and control devices located in the Control Room.
Operation and control of the reactor are handled entirely from the Control Room.
Twenty feet of space and a wall of 12 inches of masonry or 8 inches of water be-
tween
the reactor and the Control Room (Fig. 19) provide protection for the
operator from stray radiation. Viewing advantage is furnished the operator, and
some added protection from stray radiation, by the floor of the Control Room be-
ing
at an elevation of 5 feet above that of the Reactor Room.

Within the Control Room, there are two primary assemblies having to do
with operation and control of the reactor: The Rack of Data Recorders and the
Control Console (Fig. 19). In normal operating position, the operator sits at the
central panel of the Control Console, with the water-window to the Reactor Room
to his right. Directly in front of the operator, above and about 5 feet beyond the
central panel of the Control Console, the 12 data recorders are in full view. The
six central recorders, which handle data of primary importance, have illuminated
scales. Along the wall to the left of the operator are instruments and recorders
which indicate and collect data of secondary importance to the operation of the
reactor.

2. Fission chambers.

The fission chamber sensing elements of the instrumentation system are
similar in design to chambers successfully used with the homogeneous reactor at
Los Alamos.d1 Three thin concentric cylindrical shells of lucite are covered with
0.002" aluminum foils which, in turn, are coated with a layer of U235 (0.35 Mg/cm2)
of above 90% isotopic purity. The rate of fissioning is proportional to the neutron
flux from the reactor, and hence the current from the fission chamber is directly
related to the power level of reactor operation.

It is anticipated, on the basis of Los Alamos experience with similar cham-
bers
, that a current of 1 milliampere can be obtained when these chambers are in
a neutron flux of 1010 neutrons/cm2/sec.

3. Neutron Flux Measurements on Linear Scales.

In the reactor instrumentation system, there are three independent
channels having responses linearly related to the neutron flux. (Fig. 20).

(a) In the first linear channel, the fission chamber is connected directly,
without amplification, to a galvanometer located on the central operating panel of
the Control Console. The galvanometer sensitivity is matched to the fission cham-
ber
so that operation at 10 Kw produces full scale deflection on the least sensitive


[page 17]

range, and operation at 1 watt level produces full scale deflection on the most
sensitive range. There are 8 intermediate ranges.

Four meters, in series with the galvanometer and with each other, are lo-
cated
on alternate faces of the reactor shield respectively, so that the level of
reactor operation is visible to persons at any location in the Reactor Room.
These four meters have only one sensitivity and indicate 10 Kw at full scale.

(b) In the second channel with linear response, the fission chamber current
is fed into a preamplifier and then into a Brown Recorder. The level of radiation
flux is continually recorded. The sensitivity range of the recorder at any time is
identical to that being displayed simultaneously on the galvanometer of channel (a).

(c) The third channel with linear response is used in automatic control of
the Reactor. The output of the fission chamber is amplified and balanced against
the reference voltage from a Rubicon potentiometer. The latter voltage can be
controlled at will by the operator. The difference between these two signals, if
any, is amplified and applied to the motor which controls the motion of the reactor
control rod. The motion of the motor is always in the direction which would de-
crease
the difference between the signals. For the benefit of the operator, this
difference in signals is also displayed on a small cathode ray tube mounted on
the Control Console.

There are certain safety features common to each of the above systems. In
each system, a safety interlock is provided such that, in case the neutron flux
exceeds a pre-set level, the current to the electromagnets supporting the control
and safety rods is turned off and the rods are dropped. In addition, between
channels (b) and (c) there is a signal comparator which also causes the rods to
drop in case the initially matched signals differ at any subsequent time by a pre-
set
amount. This, precaution insures that one or the other of the systems does
not fail without knowledge of the operator.


[page 18]

4. Neutron Flux Measurements on Logarithmic Scales.

It is necessary to provide for the operator a set of instruments in which
the flux over a very wide range of values can be shown. This is accomplished in
duplicate channels by feeding the output current of ionization chambers into ampli-
fiers
, the outputs of which are proportional to the logarithm of the input currents.
The logarithmic characteristic of a diode is used to obtain this logarithmic res-
ponse
. (Fig. 21).

(a) In the first channel, the ionization chamber is logarithmically amplified
and is then recorded on a Brown Recorder Safety features are described below.

(b) The second channel is an exact duplicate of the first, except the current
is not recorded.

In either of these channels, if the neutron level exceeds a pre-set value,
the current to the control and safety rods is interrupted and the rods are dropped.
In addition, between these two channels there is a signal comparator which also
causes the rods to drop in case the initially matched signals differ at any subse-
quent
time by a pre-set amount.

The currents from these two channels provide the input signals to the two
respective channels described below.

5. Rate of Change of Neutron Flux: Period Measurement.

The "period" of a nuclear reaction is the time required for the power
level to increase or decrease by a factor of e. It is very important that the
op- erator know the rate at which the power level is changing, i.e., the length of the
period.

The logarithmically amplified currents from the two channels described
above are fed into parallel electronic circuits respectively, where these logarith-
mic
currents are differentiated with respect to time, to furnish indications of the
period of the reactor.

(a) The period as measured by the first of these channels is indicated and
recorded on a Brown Recorder. Safety features are described below.

(b) The second channel is identical to the first, except the output is not
recorded.

In either of these two period-measuring circuits, if the period becomes


[page 19]

shorter than a pre-set value, a thyratron interlock causes the safety and control
rods to drop. In addition, between these two period measuring circuits, there is
a signal comparator which also causes the rods to drop in case the initially
matched signals differ at any subsequent time by a pre-set value.

A Master Range Changing Switch is provided for the logarithmic and period
channels which permit simultaneous change from one range to another on these two
systems.

6. Gamma Compensated Channel.

In the U235 fission and boron ionization chambers used in the channels
described above an indeterminate proportion of the current output is caused by
ionization of the gas in the chamber by gamma rays from the reactor. The neutron
fluxes indicated are therefore too high. To provide information on the magnitude of
this effect, a gamma-compensated channel is included in the reactor instrument
system. This channel consists of two identical "fission" chambers, except one
chamber contains no U235. The difference in the signals from these two chambers,
which is displayed on the Control Console, is a measure of the neutron flux essen-
tially
independent of the gamma ray ionization in the chamber. This system furnishes
information only, and is not connected to automatic safety mechanisms of the reactor.

7. Measure of the Gamma Ray Flux from the Reactor.

While a nuclear chain reaction is in progress in the reactor, the neutron
detectors described above furnish adequate information on the level of radiation be-
ing
produced. When the chain reaction ceases, neutrons are no longer released by
fission and the neutron detectors indicate cessation of activity. The gamma radia-
tion
from the reactor, however, which results from activity of the fission products,
continues at a relatively high level. To follow the level of this activity after shut-
down
, and during operation as well, a gamma ray monitor is included in the instru-
mentation
system. This monitor furnishes information only, and is not normally
connected to automatic safety mechanisms of the reactor.

This channel consists of an ionization chamber feeding through an amplifier
to a recorder.

8. Control Console and Recorder Rack.

The Control Console consists of a 3-section desk, each section being 24"
wide and set at an angle of 135o to its adjacent section (Fig. 19). The top of each
section slopes toward the operator position at an angle of 30o to the horizontal.
The operator is thus able to view the apparatus on each panel with maximum con-
venience and minimum parallax.

The central panel contains apparatus of primary importance in the operation
of the reactor. Included on this panel are:


[page 20]

The right hand panel contains the adjustment dials and galvanometer of the
Rubicon potentiometer used in automatic operation, the shim rod positioning
switches, shim rod extreme position indicating lights, and various meters indi-
cating
auxiliary operation data.

The left hand panel contains pairs of red and green lights (red indicating
non-operating and green indicating normal operating condition) for numerous
auxiliary instrumentation systems; e.g.:

Stack radiation monitor Gas recombiner flow meter
Campus radiation monitors Recombiner coolant water flow meter
Coolant radiation monitor Reactor coolant flow meter
Stack flow meter 

Also on the left hand panel are located the switches for trip-testing the
automatic safety mechanisms in the neutron measuring channels, the meters
indicating the recombiner gas circulation rate and the recombiner coolant tem-
perature
.

The Recorder Rack contains 12 Brown Recorders. The data recorded on
each are as follows: (Single channel, curve-drawing instruments used unless
otherwise indicated.)


[page 21]

9. Control Safety and Shim Rods.

Two vertical boron rods moving inside of sheath tubes which are re-entrant
into the Reactor through the top surface, and two vertical cadmium strips located
on the external periphery of the reactor cylinder, provide the means of controlling
and adjusting the level of reactor operation (Fig. 22).

The rods are of thin-walled stainless steel tubing filled with boron powder.
The two rods and their remotely controlled actuating accessories are completely
independent, but are identical in construction and in relative reactor location, so
that the rods may be used interchangeably to perform respective functions as
Safety and Control Rods. When fully lowered, the rods extend (inside of their
re-entrant tubes) 9-1/2 inches into the reactor solution, to within 1/4 inch of the
bottom of the reactor. When fully raised, the lower ends of the rods are 1-1/2
inches above the top surface of the reactor. The length of travel is 12 inches.
With the motor at maximum speed, the rate of rod movement is 5 inches/minute.
This may be controlled at any lower rate desired.

The boron rods are connected to their respective motor-driven elevating
screws by a direct current electromagnetic coupling. If any one of the numer-
ous
potential safety signals interrupt the current to the electromagnets, the
suspended rods are released and they fall by their own weight into the reactor.
A shock-absorber slows the rate of fall over the last 1 inch of travel. The
electronic safety circuits pre-set to release the rods and the electric circuit to
the magnet are so adjusted that the rods are completely released by the magnet
in 0.02 seconds after the initial signal appears, if the reactor is operating above
the 100 watt level. At lower levels, the time of release is somewhat longer.

The two boron rods are so connected by a micro-switch interlock that one
rod must be completely poised at its upper limit of travel before the other rod
can be raised. The excess reactivity in the reactor is so adjusted that either of
these two rods alone, completely inserted, will stop the chain reaction. Thus
one rod is always "cocked" in Safety Rod position before reactor operation can


[page 22]

be initiated by withdrawl of the second Control Rod.

The boron rods are located on a diameter of the reactor cylinder on alternate
sides of the center, each 2-11/16 inches (on center) from the center, with their
axes parallel to that of the reactor cylinder. The rods are thus at position of ap-
proximately
maximum effectiveness. By calculation of the relative effectiveness
of these rods, as compared to similar rods similarly placed in the spherical Los
Alamos
Homogeneous Reactore1, it is estimated that either of the rods in the
Raleigh Research Reactor is "worth" 80 grams of U235. In actual operation, the
excess uranium, above that required to produce criticality, will always be less
than this amount.

The Shim "rods", two in number, and likewise independent in operation
and identical in construction, consist of 4" wide strips of 1/32" Cadmium, 10"
long, are located on the periphery of the reactor cylinder in a vertical position
(Fig. 22). The shim rods are positioned by a variable speed, motor driven
mechanism exactly similar to that of the control rods, except no electromagnet
coupling is provided between the motor driven screw and the shim.

In lowered position, the lower ends of the cadmium shim "rods" are even
with the bottom of the reactor cylinder. In fully raised position, the lower ends
of the shims are 1 inch above the top of the reactor. The length of travel is
12 inches; the maximum rate of travel is 5 inches per minute.

The shims are positioned with their centers 5-3/8 inches apart, on one
side of the diameter on which the control and safety rods are located, and
symmetrically spaced with respect to the Control-Safety Rod diameter. The
shims thus are each 64o from the Control and Safety Rods, and 52o from each
other. The Control-Safety Rod diameter is perpendicular to the direction of the
thermal column from the reactor, and the shims are on the opposite side of the
Control-Safety diameter from the thermal column. The shims are thus in posi-
tion
(a) approximately maximum effectiveness, and (b) minimum effect on the
flux entering the thermal column.

It is calculated, again, by comparison with the somewhat similar shims of
the Los Alamos Reactor, that each of the shims of the Raleigh Research Reactor
is "worth" 25 grams of uranium.

The shim rods are intended essentially to "normalize" the excess reactivity
of the reactor so that the control rod operates at its position of maximum sensi-
tivity
. The procedure for use is on this wise: the safety rod is raised; the con-
trol
rod is (partially) withdrawn to the position approximately desired; then one
or both shim rods (in succession) are withdrawn until a sustained reaction begins.


[page 23]

The reaction is thus achieved with the control rod at the desired position.

The shim rods are necessary because the excess reactivity of the reactor may
change from day to day as exposure samples are inserted or withdrawn, as reflec-
tor
in exposure ports is removed or returned, etc. If the shim rods were not used,
frequent additions or removals of nuclear fuel might be necessary to keep the excess
reactivity adjusted to safe and workable limits.

10. Campus Monitoring.

It is not anticipated that enough radioactivity will be released from the
reactor facility to cause significant increase in the normal level of background
activity. To insure positively that this is the case, a series of instruments designed
to measure continuously the level of radioactivity at various locations on the campus
are provided. Part of this system was placed into operation in September, 1951, in
order that a continuous record of normal background radiation might be obtained
over a period of several months before the reactor is placed in operation.

Beta and gamma monitoring stations are located at 5 positions on the campus.
A mobile monitoring station will be used in making periodic check of the area. The
stationary positions are in a deliberately chosen pattern with respect to the reactor
location: Four positions are each about 550 feet from the reactor, in northeast,
northwest, southwest, and southeast directions, respectively. The fifth station is
1200 feet northwest of the reactor. The prevailing wind in this area is toward the
northwest; therefore there are two monitoring stations in the direction of the pre-
vailing
winds and one each in directions opposite and perpendicular to this, res-
pectively
.

Two types of monitoring instruments have been chosen: A G-M tube rate-
meter
, recorder combination and an ionization-chamber, dynamic condenser, recorder
electrometer system. Two monitoring stations are equipped with one type and two
with the other. The fifth station, namely; the near station in the direction of the pre-
vailing
wids, is equipped with one of each type.

Each electrometer is provided with an automatic zero drift indiciation and an
automatic calibrating device which operate at regular intervals. The latter consists
of a clock mechanism arranged to withdraw a known radioactive source from a
shielded position and place it in a pre-determined position near the detector. A
similar calibrating mechanism is provided for the G-M units. Thus, on the recorded
chart of the background radioactivity measured by each monitor, there also appears
an hourly check of the sensitivity of the instrument and a zero recalibration if any
drift has occurred. The recording apparatus for all monitoring stations is located
in the Control Room of the Reactor Building. Signals are sent from each outlying
station to central recorders so that the operator can be informed at all times of any
change in the level of radioactivity at any of the monitoring stations.

11. Miscellaneous Monitors.

A number of radiation detecting and measuring instruments, in addition to


[page 24]

those described above, will be provided in the reactor facility. The following are
included;


[page 25]

III. REACTOR CHARACTERISTICS A. Safety Features.

When the reactor is fully assembled, at room temperature, with graphite
reflector in position, only removal of a shimor control rod) addition of nuclear
fuel, or substitution of Be or heavy water for some of the graphite reflector could
cause an increase in the reactivity. All other changes which could occur would
reduce the potential reactivity of the reactor. An increase in temperature, in-
sertion
of a non-fissionable absorber for irradiation, opening of a channel to per-
mit
escape of a neutron beam, and addition or removal of water from the reactor
are all factors which would reduce the potential reactivity of the reactor.

The decrease in reactivity as the temperature increases is one of the most
important features. There are four factors which contribute to this overall effect:
(1) As the temperature increases, the fuel solution expands, the density decreases,
and the reactivity is reduced. (2) As the temperature increases, the average
energy of the neutrons, most of which are in thermal equilibrium with the atoms
of the solution, also increases. Increasing the average neutron velocity causes a
larger percentage to be captured (without resultant fission) by the U238 atoms
present, because of the large resonant absorption cross-section in U238 at 7 ev.
This tends to reduce the reactivity, though for small temperature increases and
for high U235 enrichment this effect is small. (3) If the temperature increases
due to an increased rate of energy release by fission, the total radioactivity in the
fuel solution increases and, a proportionally larger amount of decomposition
gases--mostly H2 and 02--are released throughout the volume of the solution.
The bubbles of these gases rise rapidly to the top, but their presence in the solu-
tion
causes an overall reduction in density, and hence tends to decrease the re-
activity
. (4) As the temperature increases to a value near the boiling point of the
solution, vapor bubbles of the solution are formed. These bubbles also lower the
solution density and decrease the reactivity. A steady state temperature in excess
of the boiling point of the solution cannot be achieved, since the pressure is main-
tained
essentially at atmospheric pressure. The negative temperature coefficient
observed for the Los Alamos Homogeneous Reactorf1 varies from 0.7 gm U235/oC
at low temperature and low power to about 1.25 gm U235/oC at higher powers and at
temperatures near the, boiling point. The temperature coefficient for the Raleigh
Research Reactor is expected to have similar values.

As water is removed from the reactor, by evaporation, entrainment, or de-
composition
, two effects occur which tend to lower the reactivity: (1) the ratio
of hydrogen atoms to uranium atoms, initially adjusted to optimum, deviates from


[page 26]

optimum, decreasing the reactivity. In fact, whether water is added or removed,
the reactivity tends to decrease. The optimum H/U235 ratio is not a sharply de-
fined
number, however, may vary over a relatively large range of values, say
from H/U235 = 300 to H/U235 = 500, without great effect on reactivity. Hence,
rather large amounts of water must be added or removed before this factor de-
creases
reactivity by a significant amount. (2) As water is removed, the dissolved
uranyl sulfate tends to precipitate out of solution. This changes the homogeneous
distribution of U235 in the solution, and any such change causes a decrease in re-
activity
.

A change in the acidity of the fuel solution resulting from radiation decom-
position
of some of the acid molecules in the solution, is more likely to cause
precipitation than is the removal of a small amount of water. Hence, the acidity
of the solution is checked frequently and adjusted as necessary.

B. Reactivity and Nuclear Behavior.

It is estimated that 715 grams of U235 at 90% isotopic purity are required
for criticality at 20oC. About 770 grams are required at 80oC, the normal op-
erating
temperature. (Both figures with all control, safety, and shim rods re-
moved
.) The reactor will contain 790 grams. Thus, there is in the reactor 75
grams in excess of the critical amount at room temperature and 20 grams excess
at operating temperature. The control, and safety rods are each "worth" 80
grams of U235, and the shims 25 grams each: a total of 210 grams. Thus either
control rod alone can control the total excess reactivity with the reactor at room
temperature, and any shim or control rod alone can absorb the excess with the
reactor at operating temperature.

It is of interest to explore the behavior of the reactor if all interlocks and
instrumental safety devices should fail as the control rod is being removed.
Consider first the case in which the reactor is at room temperature and the rod
is removed slowly, so that nuclear and thermal equilibrium are maintained.

Case 1. Slow removal of control rod.

As the rod is removed, the nuclear reaction begins, heat is released, and
the solution temperature increases. The negative temperature coefficient tends
to reduce reactivity, but as the rod continues to be removed, this effect is over-
ridden
, and the temperature continues to increase. Some of the heat released
will be removed by the cooling system of the reactor, but if sufficient excess
uranium is present, the amount of heat released as the rod removal continues
will eventually be more than the cooling system can remove, and the tempera-
ture
will increase.

If enough excess uranium is present in the reactor to permit the reaction


[page 27]

to continue, despite the negative temperature coefficient, at the boiling point of
the solution, then, as the rod is removed, the temperature will increase until the
liquid boils. If the rod is removed further, more vigorous boiling will insue.
Eventually, due to "evaporation" of water or decomposition of acid, the reaction
would cease. Vigorous boiling would probably result in large fluctuations in re-
activity
as vapor bubbles formed and collapsed.

If the excess uranium present in the reactor is not sufficient to over-ride
the negative temperature coefficient up to the boiling point, as will be the usual
situation, then, as removal of the rod causes release of more heat than the cool-
ing
system can remove, the temperature will increase to some value, at which
the k of the reactor drops below 1. As the reactor cools slightly, the reaction
recommences, the temperature increases and k again drops below 1. Thus,
the temperature oscillates around the critical value with reactor operating at
such level that the rate of heat released is just equal to that removed.

It is thus clear that the steady state amount of power released by the re-
actor
is directly related to the rate at which heat can be removed. It is estimated
that about 0.5 Kw of heat can be dissipated from the Raleigh Research Reactor by
conductivity and radiation into the reflector, with no auxiliary cooling. The cool-
ing
system described in Section II, F is designed to remove about 10 Kw of heat
when the reactor is at 80oC.

Case 2. Fast Removal of the Rods.

If the rods and shims are removed instantaneously from the reactor, while
the temperature is at 20oC; an excess reactivity of 0.016 results. With this ex-
cess
, the reactor will flash to a power peak of about 170 megawatts, with an
e-folding time of about 0.02 seconds (and a total elapsed time of about 3-1/2
seconds) before the negative temperature coefficient can begin to control the
power downward. Another 1-1/2 seconds are required to reduce the excess re-
activity
to zero.

In making these estimates, a constant negative temperature coefficient of
2.0 x 10-4/oC is assumed, and no consideration is given to the effect of bubble
formation in the solution. It is known that bubbles will be formed, and that more
favorable temperature coefficients will exist over a portion of the event. Both
of these factors would tend to slow the rate of rise, lower the value of peak
power achieved, and hasten the return of the power level to nominal values.

The net results of an event of this kind would be (1) the temperature of the
solution would be increased, probably to boiling, and (2) the levels at the outer
surface of the reactor shield would flash to a few thousand times continuous ex-
posure
tolerance levels for about 3 seconds. An explosion would almost cer-
tainly
not occur.

If all the rods and shims are removed from the cold reactor (20oC) at a
uniform rate in a total of 5 second, (the normal time required is 2 minutes) a
flash up to a power level of 14 megawatts results. (Fig. 23). As above, a


[page 28]

conservative value of the temperature coefficient and no effects of bubbling are
assumed. The flash up and return to normal levels again occurs within three or
four seconds. As shown earlier, with slow rate of rod removal (i.e., in times
longer than 15 or 20 seconds) the power rises smoothly to equilibrium value with
no flash up.

C. Power Level, Radiation Fluxes.

It is anticipated that the normal maximum power level of operation of the
reactor will be about 10 Kw. Calculations have been made of the radiation fluxes
to be anticipated at various points in and around the core, when the reactor is op-
erating
at this level. The calculated values are shown in Table 5.

TABLE 5 - CALCULATED RADIATION FLUXES AT 10 Kw. OPERATION

 Fast n Slow n Gamma
 (n/cm2/sec)(n/cm2/sec)([gamma]/cm2/sec)
Inside Core 2 x 10115 x 10116.7 x 1011
Surface of Reactor Envelope 7 x 10103 x 10114.5 x 1011
Surface of Reflector (20" from
reactor)
 8 x 1098 x 10101.8 x 1010
Outside 4" lead shielding ""1.1 x 108
Four feet from center of
reactor on axis of thermal
column
 2.5 x 1071.4 x 10106.0 x 107
One foot from end of thermal
column (on axis)
 2.0 x 1051.5 x 1091.2 x 106
Outer surface of concrete at
end of column
 70
Average radiation over outer
surface of concrete shield
 0.8negligible0.15


[page 29]

D. Fission Products; Radioactivity.

After the reactor has been in operation for some time, the solution will
possess a considerable amount of radioactivity, due to the build up of fission pro-
ducts
. It is necessary to know how much activity accumulates in the solution and
how much radioactive gas is evolved.

If the reactor were operated for a known time at a known power level, cal-
culation
of the resulting fission products would be a relatively simple matter.
For intermittent operation at random levels and for various lengths of times,
calculation of the composite accumulation of non-volatile and gaseous products
becomes impossible. To determine the upper limit of activity to be expected,
a schedule of operation was postulated which results in a build up of activity not
expected to be exceeded by any actual schedule of reactor operation. The pos-
tulated
"worst case" of reactor operation is:

(1) The reactor is operated at 10 Kw for the first 6 hours of every suc-
cessive
24 hour period, and is not operated over the remaining 18 hours of the
period. (At times, the reactor may be operated continuously for periods con-
siderably
longer than 6 hours. Such periods would be preceded and followed
by correspondingly longer "down" periods, however, so that the average re-
sult
, say at the end of' the week, would produce no more activity than that
produced by the schedule described.)

(2) Assume that the gas disposal system operates as described in Section
II G for the 6 hours that the reactor is in operation, plus 1 additional hour after
shut down, and then does not operate over the remaining 17 hours of each 24
hour period. The' gases which accumulate in the reactor during the 17 hour
"down" period, however, are withdrawn during the subsequent operating period
along with those produced during the period.

(3) The refrigerated condenser of the gas disposal system lowers the gas
temperature to 13oC, and only those fission products remaining volatile at this
and lower temperatures escape from the reactor.

Two questions are asked: After the reactor has operated for many days
on this schedule, so that equilibrium has been reached, (a.) what is the level of
the radioactivity in the reactor solution, and (b) what is the quantity of radio-
active
gas to be disposed of in the 7 hour operating period of each 24 hour
period?

(a) Radioactivity of the Fuel Solution.

According to information presented in an article by K. Way,a8 the amount


[page 30]

of radioactive energy released by the fission products in a reactor immediately
after shutdown, after the reactor had been operating for an extended period at a
10 Kw power level would be 2.6 x 1015 mev/sec. It is difficult to interpret
this in terms of curies. A rough approximation may be made by assuming that
one beta and one gamma, each of 1 mev energy are given off in each "disintegra-
tion
". There are then 1.3 x 1015 disintegration/see; or 35,000 curies

It may be shown by rather simple calculations that the fission product activity
of shutdown in a reactor operating on a regular schedule of 6 hours per day at a
10 Kw level is about 75% of the activity at shutdown in a reactor operated contin-
uously
at a 10 Kw level. Thus the maximum activity expected in the Raleigh
Reactor is about 2. x 1015 mev/sec., or about 26,000 curies. The decay in
activity after shutdown is rapid. After 18 hours, the normal down time in the
postulated daily schedule, the activity would be only about 2.6 curies.

It is of interest to note the thermal effects of the fission product activity
after the reactor is shut down. If the reactor cooling water should cease to
circulate when the reactor should be closed down, low much increase in temper-
ature
might be expected?

For data presented in the above mentioned article,a8 it may be shown that
the thermal energy resulting from the beta and gamma activity of fission pro-
ducts
in a 10 Kw reactor after shut down is given approximately by the equation:

where t is the time in seconds after shutdown, and To is the time
in seconds the reactor had operated prior to shutdown.

By application of Way's relation for activities at various times after shut-
down
, and assuming that all betas and gammas are completely absorbed in the
reactor, the rate at which thermal energy is imparted to the fuel is

If the reactor should be at 80oC, 500 watts for about 15 minutes (assum-
ing
no heat loss) would be required to raise the temperature to boiling. It is
estimated, on the other hand, that about 500 watts of power will be lost from the
reactor by condensation conduction when the temperature is at 80oC. Therefore, heat
from the fission products would hardly cause any increase in temperature, and
certainly would not cause the solution to boil.

(b) Daily Gaseous Radioactivity Produced.

Most of the primary and the daughter elements produced in the fission


[page 31]

process are neither gases nor vapors at the temperatures prevailing in the reactor.
A few, however, are sufficiently volatile to escape from the 80oC fuel solution in
the reactor. A careful examination of the, list of product elements leads to the con-
clusion
that two such elements, iodine and Xe133, the daughter of I133, contain
many times more activity than all others. Without the use of the refrigerated con-
denser
in the Recirculation-Recombination system,(Section II G), it is possible that
a considerable portion of the iodine might find its way out of the reactor solution,
through the disposal system, and out into the atmosphere. This would probably
necessitate a revision of the disposal system, for the tolerance concentration of
radioactive iodine in the atmosphere is very low. (10-7 micro curies/cc)
Available evidence indicates, however, that the iodine will be returned to the re-
actor
by the refrigerated condenser, even if it escapes into the Recombination
system.

This leaves the daughter product, Xe133, to be handled in the gas disposal
system, for all xenon produced would be expected to escape from the reactor. The
calculation of the amount of Xe133 produced by the intermittent reactor operation
schedule described above is performed by essentially standard methods. The
growth-decay equation

is solved for the number of atoms of the principal primary source (I133, having a
22 hr. half-life) resulting from a single 6 hour operating period. Here g, the rate
of generation by fission, is proportional to the power level. The contribution of
all previous days is added by a series method, with the effect of decay accounted
for, to yield the equilibrium level of iodine. The gaseous Xe133 (5.3 day half-life)
that is produced by the iodine over a 24 hour period may then be determined. No
account is taken of the distribution of fission products by atomic number, and the
predecessors of I are taken to have effectively zero half-life. An outline of the
calculation is given below. For a given operating period, say from time zero to
time [tau], the solution of the above equation is

For the down-time between [tau] and 1 day, the equation reduces to
, having a solution,

Since the contribution to the number of previous days' operation is given by the latter
expression with time t - [tau] + 1, t - [tau] + 2, etc. inserted, a geometric series


[page 32]

maybe written. This is summed by the relation l + x + x2 + ---- = 1/1-x.

During the period of reactor operation, an additional growth term must be added
to this solution.

The equilibrium iodine level is different from that for continuous operation
N1 = g/[lambda]1 by a factor , with relatively small periodic varia-
tions
reflecting the nature of the operation. The Xe accumulation per day is then
derived from the solution of the equation

where the last term is the fluctuating supply from iodine decay.

For the chosen pattern of reactor operation, i.e., 6 hours on and 18 hours
off in each 24, the equilibrium level of I133 is computed to be around 365 curies,
and, hence, the daily Xe production is 11 curies. To reduce the 11 curies of
Xenon activity to a value below 1 curie, a hold up of about 4 half-lives, or 21 days,
is necessary. The system designed to accomplish the desired hold-up of the Xenon
actually achieves reduction of the activity to a value many times lower than this.
In this system, the gas from the recombination unit is bubbled through a water
trap to the first of several holding tanks arranged in series. A water trap is lo-
cated
between each tank so that the gas bubbles from one tank to another through
these traps. The delay of the gas in reaching the stack after traversing this system
of hold-up tanks is calculated according to the analysis below.

Three types of holding systems are considered: (1) a single long continuous
tube, (2) a single large tank, (3) a sequence of small tanks.

Since the effectiveness of a given holding volume can be shown to be dependent
on the degree with which mixing is prevented, it follows that the long tube is the
most favorable system, and the large tank is the least favorable. The analysis of
the relative merits of the three arrangements was made on the basis of continuous
flow; the extension to intermittent flow introduces only a correction factor.

Case 1. Regardless of the dimensions of the system, the time t for a given
sample of a fluid flowing without turbulence or friction through a vessel is given by
V/v where V is the total volume of the vessel and v is the volume flow rate. The
radioactive attenuation factor is thus

where [lambda] is the decay constant and tH is the half-life of the radioactive consti-
tuent
. This situation would most nearly be approached by a long tube.


[page 33]

Case 2. The rate at which molecules (radioactive or not) will be exhausted
from the large tank is given by

Letting v/V = f, the solution if the initial number introduced at time zero is N0,
N = N0e-ft. Complete mixing of the contents are assumed. The total of the
particles that escape as radioactive over all subsequent time is given by

so that the attenuation factor is .

Case 3. If the volume V is made up of n tanks of volume V/n, the above form
for the attenuation factor may be applied successively, giving the result

(By the mathematical definition of the logarithmic base e
it may be shown that as n approaches ∞, the formula
for Case 1 is obtained, as would be expected from physi-
cal
considerations.)

Complete mixing is again assumed.

The expected value of the discharge rate is 100 ml/mm for 6 hours, giving 36
liters. Let the collected air be released continuously at a rate v of 36 liters/day.
Let the total volume V be 800 gallons or 3028 liters. A single tank would reduce
the Xe activity [lambda]= 0. 131 days-l) by a factor

The single pipe of the same volume would reduce the activity by

The length of say 1/4" ID tubing needed to contain 800 gallons is prohibitively
long however, being of the order of miles.


[page 34]

Case 3. A set of 8 tanks each of 100 gallon volume should give an attenuation
of

The evaluation of the system of tanks with intermittent operation is obtained
by the solution of the differential equation for the concentration of radioactive ma-
terial
, including the effects of decay and escape. The contribution of one days'
operation is found from the relations
during operation 0 < t < [tau]
during shut-down [tau] < t < 1 day

where C is the variable tank and discharge concentration and C0 is the inlet
concentrates.

The solutions are
0 < t < [tau]
0 < t < 1 day

The maximum concentration resulting from many days operation is given by the
sum

Since the discharge occurs only during the time [tau] = 0.25 day, the flow
rate v is four times that in the previous calculation, i.e., 144 liters per day.

Substitution of other appropriate numbers for the case of 8 tanks yields
Cmax = (0.441)Co


[page 35]

It will be conservative to assume that this maximum serves as feed for the
second tank, in which a further attenuation of 0.442 will occur, etc.

Thus, by extension

The main precaution necessary to guarantee the applicability of this analysis
is that "streaming" of freshly generated material through the system be avoided.

With a system of 8 one hundred gallon tanks in series, therefore, an attenuation
of the Xenon activity by a factor of 700 is expected. The 11 curies of activity is re-
duced
to less than 0.02 curies.

A number of fission product gases other than Xenon may be produced. How-
ever
, a careful survey has failed to reveal any other which is produced in sufficient
quantity or with sufficient activity to be as hazardous as the Xenon. All others are
produced in small relative amounts, and their half lives are either as short or shorter
than that of Xenon, in which case their attenuation would be comparable to that of
Xenon, or their half lives are quite long, in which case the activity is negligibly low.

The activity of the gases released daily from the reactor system, including
Xenon and all others, is therefore expected to be well below one curie.

If, however, the daily amount released should greatly exceed expectations,
say to the total of 1.5 curies over the 7 hour operating period, no hazard would en-
sue
, An activity of 1.5 curies, diluted uniformly into the 12,500 cfm stack stream
and dispersed into the atmosphere over the operating period, would result in a con-
centration
of only 1 x 10-5 microcuries/ml at the stack discharge. For the type of
material (permanent gases) and the low total amount involved, this is considered
within the permissible concentration level.


[page 36]

IV - REACTOR HAZARDS AND SAFETY PRECAUTIONS A. Normal Hazards.

In the routine operation of the reactor and its associated laboratories cer-
tain
hazards to personnel will exist. The situation is analogous to that existing
in an X-ray laboratory or in a chemicals manufacturing plant where toxic gases,
say, fluorine for example, are handled. In these and all similar situations,
safety to 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 the reactor facility are listed below,
together with the means of insuring safety of personnel.

1. Radiation.

Radiation may come from two sources: (1) The Reactor, and (2) Radio-
active
sources, e.g., irradiated samples.

If a reactor port were open and a direct beam, say 1 or 2 inches in diame-
ter
, were allowed to emerge, with the reactor in full operation, a dosage of
several thousand Roentgens per minute in the beam and considerable quantities of
scattered radiation over the entire room would result. No actual hazard to per-
sonnel
from this source is anticipated, however, for several reasons. It is not
intended that personnel will be present in the Reactor Room when the reactor is
in operation, except in carefully pre-arranged cases. When experiments necessi-
tate
a direct beam from the reactor outside of the reactor shielding, very special
precaution will be exercised to keep personnel out of the beam path. Physical
barriers will be erected, and adequate thicknesses of shielding along the beam
path to reduce the amount of scattered radiation will be used.

In the wall of the reactor room at a point opposite each beam port where a
direct beam from the port would strike, a 15" diameter hole leading to a 10 foot
radiation trap buried in the earth under the laboratory is provided. The amount
of radiation scattered into the room is thus minimized.

To guard against radiation which might leak through crevices in the exposure
port plugs and from other radioactive sources, the following precautions will be ex-
ercised
:

(a) All direct beam ports are at a height of 24 inches from the floor so that
exposure due to stray leakage from the ports would occur on the legs of personnel.

(b) The beam ports are provided with offsets along their lengths, so that no
straight line escape paths for the radiation is left between the walls of the port and
the plugs that normally fill the ports.

(c) The walls of the reactor room are constructed of 12" masonry to prevent


[page 37]

scattered radiation reaching personnel in other parts of the building. In the Obser-
vation
and Control Rooms, water windows 8" thick permit observation without ex-
posure
to stray radiation.

(d) For holding stored radioactive materials, a bank of tubes project at a
downward angle into the earth from a height of 4 - 6 feet at several places in the
wail of the reactor room. Thus, persons in the reactor room are protected from
radiation from these stored materials.

(e) Radiation monitors, equipped with visible and audible warning signals,
are placed in strategic positions in the building to provide warning when radiation
tolerance limits are being approached. Warning lights on the reactor and in var-
ious
parts of the building indicate when the reactor is in operation.

(f) All personnel in the building will be required to wear personnel electro-
scope
monitors, film badges, or other radiation recording devices as necessary,
to provide knowledge of routine exposure history.

(g) Constant vigilance and emphasis on safety will be demanded of all per-
ons
engaged in activities in the vicinity.

2. Radiochemical and Radiophysical Hazards.

In the pre- and post-exposure handling of samples and specimens, con-
siderable
manipulation of radioactive materials, largely beta and gamma
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.

(a) No person will be permitted to engage in radiophysical and radiochemical
experiments who has not satisfactorily demonstrated adequate training to perform
the proposed experiments.

(b) An adequate supply of radioactivity instruments and equipment, including
all necessary protective equipment, will be available. Ventilation will provide
motion of air from areas of lower activity toward those of higher activity. In hoods
and on chemical benches, ventilators will move air away from the operator. Ade-
quate
tools, tongs, shielding brick, and radiation monitors will be provided in the
laboratories.

(c) No smoking or eating in radioactive areas will be permitted.

(d) Lockers, showers, clothing change rooms, scrubbing facilities; hand,
foot, and clothing monitors will be provided and their use will be required.


[page 38]

B. Minor Incidents.

In this category are listed inadvertent, unexpected, unplanned and abnormal
occurrences 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
will be released. The reactor envelope is provided for this occurrence. The
released liquid and gas will be contained in the envelope. The Liquid Salvage Line
(Section II, d) and a vacuum pump may be used to withdraw the escaped liquid
from the reactor envelope into shielded containers of safe geometry. The gases
may be pumped from the envelope by purge lines to absorption traps or to a gas
holding tank and subsequently 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 "hot" apparatus. If the parts to be repaired cannot be de-
contaminated
; i.e., if the metal itself is highly active, radioactive decay must be
awaited or the part must be discarded.

In case of leakage of the fuel solution from both the reactor and the reactor
envelope, small drain holes in the graphite blocks immediately below the reactor
permit the fluid to flow freely to the secondary stainless steel catch basin
(Fig. 12) below. Calculations show that a nuclear chain reaction would not occur
if all the fuel in the reactor were distributed by leakage into the graphite under the
reactor

2. H2 - O2 Explosion.

The hydrogen and oxygen present in the reactor resulting from the radia-
tion
decomposition of the fuel solution offer the possibility of explosion unless
sufficiently diluted. It is believed that the Recirculating-Recombination apparatus
described in Section II, G, will effectively eliminate all explosive hazard by keep-
ing
the combustible gas at a dilution below the explosive limit. Should an explosion
occur, however, it could cause damage or rupture of the reactor or the gas disposal
system.

A quite violent H2 - O2 explosion would be required to cause damage to the
Reactor Safety Envelope as well, because of the relatively large volume and the
geometrical arrangement of the latter. In case such a violent explosion should
occur, in which the Reactor Envelope should be damaged, the resultant release of
radioactive gas would constitute a major catastrophe. The hazards associated
with such an event are discussed in Section C, below.

In case a H2 - 02 reaction of lesser violence should occur, in which no
damage to the Reactor Envelope is involved, the hazards entailed would be no
more serious than those associated with a leak in the reactor, and the protective


[page 39]

and remedial measures would be as described above for an event of this sort.

3. Failure of Water Supply.

If a failure in the water supply should occur, two results might follow;
neither would entail a hazardous consequence. One result would ensue in case the
failure were such that the water merely stopped circulating with the coils remain-
ing
full. At any given setting of the control rods, if circulation of the reactor
cooling water should cease, the temperature would increase and the power level
of the reactor would decrease. This decrease would continue until the rate of heat
release should just match the rate of heat dissipation by conduction through surfaces
of the reactor. This power level would have a value around 0.5 Kw. Thus, water
failure would result in a reduction in power level to about 0.5 Kw.

If the cooling water should cease to circulate with the reactor on automatic
control, the temperature would increase as the automatic controls moved the con-
trol
rod out in an effort to maintain the power level constant. When the tempera-
ture
increased to the pre-set trip value, the rods would drop.

The second result would ensue, in case the water failure were such that the
water were drained from the cooling coils. Removal of water from the cooling
coils, all other factors remaining constant, causes an increase in reactivity of the
reactor. If the water were removed at a rate no greater than the normal flow rate,
the reactivity would increase at about the rate indicated in Curve B of Figure 23
until the coils were empty. Thereafter the usual temperature effect would set in.

4. Escape of Radioactive Gases from the Stack.

In the laboratories of the reactor building radioactive materials up to
about 1 curie in activity may occasionally be handled. Should an accident occur in
which this amount of radioactivity were released, and if this material should be
volatile in nature, it might be picked up immediately by the ventilation system and
discharged through the stack into the atmosphere (The stack monitor would
normally turn off the blower) To evaluate the hazard which could result from
such a release, under the most unfavorable conditions likely to be encountered,
the following circumstances have been postulated:

The normal air discharge from the stack is 12,500 cubic feet per minute.
Thus, in 10 minutes, 125,000 cubic feet of air, containing 1 curie of radioactive
gas, would be released at the top of the stack. If the mixing were uniform, an
activity of 3 x 10-4 microcuries per milliliter would result. Depending on the
identity of the radioactive material, this level of activity would be from 10 to
100 or even, for one or two materials, 10,000 times the normal permissable


[page 40]

concentration (for continuous exposure).

In the most unfavorable case, this discharged volume of air would accumulate
as a sphere, 40 feet in diameter, and remain as a more or less undispersed "cloud"
as it drifted away with the wind. The potential hazard to persons in the path of this
"cloud" would depend on the type of radiation involved, the half-life of the active ma-
terial
, the wind velocity, the rate of dispersal, the elevation of the "cloud" and
finally, the likelihood of any of the material being ingested and the retention charac-
teristics
should ingestion occur. Thus, about half the variable factors relate to the
nature of the radioactive material released, and the remainder to the meteorological
conditions existing at the particular time.

Certain data from records of the Raleigh area were presented in the earlier
report on reactor design.a2 Additional data are presented in Appendix I, of this
report, and, further collection and analyses of meteorological date are in progress.
In general, it might be pointed out, that the wind blows with a velocity of 5 to 15
miles per hour for more than 80% of the time; there is calm (below 1 mph) very
seldom and wind velocities below 5 mph only 10 to 12% of the time.

The chances are quite good, therefore, that the radioactive "cloud" described
above would be carried away from the point of release at a rate of 5 to 15 miles
per hour, and that the associated dispersive tendencies would occur. A frequently
quoted rule for dispersion is: The lateral spread is one-seventh of the distance of
travel.

For the purpose of calculation, therefore, it is assumed that a wind velocity
of 5 miles per hour at the time the radioactive gas is released, and that the gases
drift downwind as a spherical volume which increases 100 feet in diameter with
each 700 feet of travel. At 1400 feet, the lower "surface" of the spherical cloud
reaches ground level. The concentration of activity has decreased to 1.5 x 10-6
microcuries/ml (assuming dilution only, no radioactive decay) at this point.

A person in the path of this radioactive cloud would receive a more or less
hazardous exposure, depending on the nature of the radioactive material. For
most materials, exposure for the few minutes required for the passage of the
cloud, at the concentration indicated, would constitute no hazard whatever. Ma-
terials
which are highly retained if ingested, such as iodine, are more dangerous.
Exposure to even these materials for the short time involved in the passage of
the cloud, and at the concentrations indicated, would cause no injury.

If the amount of activity released from stack should be 10 or 100 curies,
instead of 1, the hazards would be increased correspondingly and the possibility
of injurious exposure would be greatly increased.

5. Addition of Water of 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 water is added or removed the reactivity tends
to decrease. Thus, filling the free volumes of the reactor, i.e., above the liquid


[page 41]

surface, in the control rod scabbards, etc., would reduce the activity of the reactor.

Filling of the free volumes of the reactor with additional U235 solution would
increase the reactivity of the reactor. About 1000 cc of solution, containing at op-
timum
concentration, 62 grams of U235 could be placed in the reactor above the
normal liquid level. Eight more grams of U235 could be placed in the vertical re-
entrant
exposure port, and thirty grams inside the cooling coils (if the water were
first removed). Thus, a total of about 100 grams of U235, at optimum solution
concentration, could be placed inside the reactor's free volumes if one were de-
liberately
intent on creating the greatest possible excess reactivity. The reactor
rods and shims together would be more than adequate to control this increased
uranium content, but any one alone would not.

6. Reprocessing the Reactor Fuel Solution

If the reactor should operate at 10 Kw power level 24 hours per day and
365 days per year, about 3 grams of U235 would be "consumed". It is probable
that scarcely more than 1/2 gram per year will be used in the first several years
of operation. Thus, reprocessing of the nuclear fuel should be a very infrequent
requirement.

When reprocessing does become necessary, due to the build up of fission
product poisons, the solution will be withdrawn through the Sampling Tube
(Section II, D) into shielded containers of safe geometry and shipped by A.E.C.
approved means to a designated A.E.C. chemical processing plant for purifica-
tion
. No chemical manipulation, beyond that involved in the original filling
operation, will be necessary at the reactor site.

C. Major Catastrophes.

In this category are included those unforeseen and unplanned events and
accidents of such violent proportions that the reactor and the reactor building
may be wrecked and the lives of persons in the vicinity endangered. Only two
means are considered by which such catastrophe could occur: (1) earthquake or
other act of God, and (2) sabotage by (non-nuclear) explosion. Any violent and un-
controlled
nuclear reaction resulting in explosion is excluded since it is believed
that an event of this type could not occur.

If a catastrophe should occur, only one real hazard could ensue, beyond
that entailed by the event causing the catastrophe, namely; contamination of the
area with radioactive liquid and gases.

There are two possible types of `hazards which could result from an earth-
quake
. The first type entails abnormal nuclear behavior of the reactor due to
disturbance or dislocation of the control and safety rod mechanism. In the worst
possible case, the rods would be violently removed from the reactor, causing a
momentary power flash-up to high levels before the negative temperature coeffi-


[page 42]

cient and bubble formation bring the power, down to an equilibrium value.
(Section III, C).

An event of this type would cause no hazard to personnel unless the event
were accompanied by gross disturbance of the concrete shielding around the re-
actor
. In this case, other types of hazard as discussed below would be more
severe and prolonged.

The second possible hazardous consequence of an earthquake would involve
major damage to the shield and rupture or destruction of the fuel cylinder. The
fuel solution would be spilled out, and the contained radioactive materials, both
liquid and gaseous, would be released. The hazards involved would be similar
in type, though less in magnitude, than those described below for sabotage by
explosion. The two will be discussed together, for consideration needs to be given
at this point only to the most unfavorable case.

A carefully placed non-nuclear charge of explosive could completely wreck
the reactor core and release the fuel solution. Dispersal of the accumulated radio-
activity
contained in the reactor could cause severe damage to personnel in the
vicinity. The chief danger would come from the volatile or volatized portions of
the radioactive materials

All non-volatile materials would either spill onto the floor or be thrown upon
the ground. The chances of the liquid spilled on the ground entering an underground
flow channel and eventually reaching the city water supply are vanishingly small,
because of the nature of the topography, and the soil and subsoil formation in the
area. Any liquid spilled onto the floor would collect first in the drain sump, and
from there could only get into the city sewer system by being pumped into the
laboratory drainage system above and then flowing through the system of holding
tanks, with their automatic valves and monitors, to the city system (Section I, C).
The positive actions required to accomplish this, and the numerous automatic
safety interlocks which would act to prevent it, would almost certainly insure that
the liquid would not get into `the city sewer system.

In considering the hazards arising from release of volatile radioactive ma-
terials
, three levels of intensity or destructiveness of the causative event may be
envisioned. In the first, the reactor core itself is destroyed and all its contents
are released, but the shielding assembly is not damaged or dislocated.

This case is the one most likely to occur, for a very powerful explosion
indeed would be required to demolish the massive shield structure. If radioactive
gases should be released inside the shield, their escape into the reactor room
through the crevices and openings in the concrete would be at a relatively slow
rate. The air in the room would become highly contaminated and the gases would
be picked up by the ventilation system. There would be plenty of time, however,
and plenty of warning from monitoring instruments, to enable the operator to de-
cide
on a course of action. With the reactor room closed and the ventilating system
turned off, for example, most of the radioactive material could be held for an ex-
tended
period.


[page 43]

A catastrophe of the second level of intensity is one in which the reactor core
and the reactor shielding are destroyed or severely damaged, but the walls of the
reactor room remain intact. Again, the contamination may be picked up by the
ventilation system and blown into the atmosphere. By turning off the ventilation
system (or leaving it off) and keeping all doors into the reactor room closed,
however, the escape of the radioactivity could be greatly retarded.

In a catastrophe of the most violent proportions, destruction or severe damage
to the reactor, the shielding and the reactor building might occur. In this case a
good portion of the reactor solution could be vaporized in the catastrophic event,
and the radioactive cloud so formed would be carried from the site by the wind.

In order to calculate the hazards involved, should such an unlikely event
occur, the following assumptions are made.

1. That the reaction has been operating continuously for an extended period
at 10 Kw, and when the catastrophe occurs, the reactor, the concrete shielding, and
the reactor room walls are destroyed.

2. All of the accumulated radioactive materials in the reactor are volatized
into a spherical cloud of 500 feet in diameter.

3. The cloud is formed essentially at ground level, and drifts away at a
velocity of 3 miles/hr. A period of 10 seconds elapses after the catastrophic event
before the cloud forms and exposure of nearby personnel not involved in the
catastrophic event itself, begins.

We calculate then the hazards to personnel exposed to this cloud of radio-
activity
:

Using Way'sa8 formula for the amount of power in the fission products,
from a reactor which has operated for an extended period at 10 Kw,

where t is the time after shut-down. For example when
t = 10 seconds, P = 402 watts.

To compute the radiation dosage resulting from this amount of radioactivity,
the conversion formula given in Report Wash. 3 (of the USAEC) is used:

For a 500 foot diameter cloud, which travels at a rate of 3 miles per hour,
the maximum exposure is,
R = 12.3 roentgens


[page 44]

For persons farther away, the exposure dosage would be less than this.

Thus, even for this severe catastrophe, the radiation exposure would not
be alarmingly large. There would be added hazard, due to ingestion of radio-
activity
by a person "immersed" in the cloud, but for short periods, even this
would not be dangerously large.

A considerable number of precautions against sabotage have been incor-
porated
into the reactor design:

(1) All doors, windows and other outside openings in the building are kept
closed and locked except when authorized personnel are present.

(2) The doors to the reactor room, particularly, are provided with good
locks, which can be unlocked with a key only after electric release of the lock
from the Control Room.

(3)Electrical current to the crane is turned off and the switches are locked
except when authorized personnel are present.

(4) All external openings into the concrete shield of the reactor are closed
when not in use, by "burglar proof", combination-lock, safe doors.

(5) A daily inspection of the reactor and its control apparatus by a respon-
sible
staff member is made, e.g., on Sundays, holidays, etc., whether research
is in progress or not.

(6) The reactor building and the area about the building is kept well-lighted
at night.

(7) College watchmen and city polic direct particular attention to the re-
actor
building.


[page 45]

V. OPERATION AND EXPERIMENTAL PROGRAM A. Start-Up.

When the reactor assembly has been completed and the fuel is ready to be
charged, a neutron source (1/2 - 1 curie) will be located inside the re-entrant
sample exposure tube at the center of the fuel cylinder, and the cylinder will be
filled with water to normal liquid level. Measurements will be made of the
neutron fluxes at all instruments with this "zero reading" configuration, with all
control and shim rods completely raised. At least four independent neutron
measuring instruments will be used.

Some of the water will then be withdrawn, and an equal volume of uranyl
sulfate solution will be placed in the fuel cylinder. Again neutron flux measure-
ments
will be made. Using these data, and those to be obtained from subsequent
steps, curves of reciprocal counts versus mass of U235 will be plotted as indicated
by each neutron measuring instrument. The X-intercept of these curves indicates
the amount of U235 required, to produce criticality under the existing conditions.
It is known that these curves may not be linear, and therefore the X-intercept
will be approached with caution.

Each addition of uranium to the solution will be made with one safety rod
fully removed so that, should the extremely improbable event occur of criticality
being achieved during fuel addition, the rod could be dropped to control the excess
reactivity The other rod and the shims are fully inserted during fuel addition.
Each of the shims and the rod in succession are then slowly removed until all four
rods are fully withdrawn, The neutron flux measurements are then made.

The stepwise addition of nuclear fuel is repeated until criticality is reached.
The added increments of U235 are smaller as criticality is approached.

B. Initial Experiments.

It is anticipated that the reactor will be operated at low power, with little
excess uranium above the critical amount, for a considerable time. The charac-
teristics
of the reactor, the flux distribution, the behavior of the instruments and
of the auxiliary systems will be determined Adjustments and revisions will be
made as necessary.

As operational experience is gained, the power level will be increased, until
maximum power is reached. Enough excess uranium, above the critical amount
at operating temperature, will be added to permit some flexibility in research pro-
cedures
. It is not anticipated, however, that more excess uranium, above the
critical amount at room temperature, than can be controlled by one rod alone will
be added.


[page 46]

C. Subsequent Experiments.

It is expected that the research program associated with the reactor will
consist of three general types of experiments. (See Nucleonics, 9, No. 5,
November 1951, for a complete discussion.)a9

(a) Reactor Characteristics.

Included here are studies of temperature effects, flux distribution,
effects of voids in the core or reflector, and changes in flux by local poisons and
control rods, etc.

(b) Irradiation of samples.

(c) Use of radiations from the reactor, for experiments in physics,
chemistry, biology, and related fields.

D. Procedures and Policies.

When the reactor routine has been established, a set of procedures and
precautions will be formulated and rigidly followed in daily operation. Some of
the policies by which the operation of the reactor and its associated facilities will
be guided are listed below. These general policies will be extended into the estab-
lishment
of such specific rules and regulations as may be necessary for the safe and
efficient use of the reactor facilities.

1. Complete responsibility for operation of the reactor and its associated
facilities is allocated to a Scientific Director. The present Director is Dr. Clifford
Beck
. Designation of subsequent directors will be agreed upon by both State College
and the official representative of the Atomic Energy Commission.

2. Advisory Committees as may be needed from time to time will be selected
by the Scientific Director to assist in the establishment of policies, rules and regu-
lations
governing operations.

3. The widest and most extensive use of the Reactor will be encouraged. Its
availability for use on acceptable projects will be limited only by considerations of
safety and efficiency of operation and availability of finances.

4. Individuals or groups desiring irradiation of samples or other use of the
reactor, not involving experimentation on human beings, must produce evidence to
the satisfaction of the Scientific Director that they are competent and that their


[page 47]

facilities are adequate to handle the radioactive materials which may be involved.

5. For projects involving experimentation on human beings, prior approval
must be obtained from the Atomic Energy Commission's Isotope Branch in addi-
tion
to stipulation 4 above.

6. Careful and complete daily log records will be kept of all reactor opera-
tions
, sample irradiations, etc., and of all individuals participating in experimen-
tation
involving the reactor.

7. All proposed projects involving usage of the reactor must be submitted in
writing and be approved in advance by the Scientific Director.

8. Responsibility for operation of the Reactor may be assigned only to ex-
perienced
staff members.


[ Appendix I]

APPENDIX I

ADDITIONAL METEOROLOGICAL DATA OF THE RALEIGH AREA

Meteorological data of rather sketchy nature have been collected in the
vicinity of Raleigh for some 60 years. Data on wind directions, velocities and
precipitation have been collected for only 7 years. These latter data have been
collected at a Weather Bureau station on the State College Campus, and hence
are directly applicable to questions relative to operation of the nuclear reactor
to be erected at this location.

Three tables and three figures are presented below which indicate cer-
tain
meteorological data having relevancy to reactor operation. The most im-
portant
item not included is that of frequency and description of temperature
inversions. These data are not available. The local Weather Bureau has ad-
vised
, however, that daily temperature inversions probably occur as frequently
as 80 to 90% of the time. The inversion usually begins during the night and
continues until early or mid-morning. Procedures are now being instituted
whereby factual information on this matter will be obtained.

Table 6 contains data on the average monthly distribution of wind direc-
tions
and velocities. The wind velocity is from 5 to 15 miles per hour more than
80% of the time. For most of the months, a southwest wind is most prevelant,
and a southeast wind least prevalent.

In Table 7 are presented data on average direction and velocity of the wind
and frequency of precipitation as observed at 3 AM, 10 AM and 3 PM daily over
a 7-year period. In general, there is no average difference between these
meteorological phenomena at different hours of the day. Precipitation occurs
most frequently with a northeast wind, and least frequently with a northwest,
southeast or west wind. This pattern varies slightly from month to month
(not shown).

The data presented in Table 8 were obtained from observations at 3 AM,
10 AM, and 3 PM over a period of 7 years. Shown are the total number of times
the wind was in a certain direction and the total number of times precipitation
was observed in the months indicated, over the 7 year period. This table gives
some further elucidation, by months, of the material presented in Table 7 above.

Figures 24, 25, and 26 present in graphical form some of the data con-
tained
in the tables above. The diagrams show wind direction, velocity and
percentage frequency, month by month, obtained from averaged data taken over
the years 1945 - 1948 inclusive.


[ Table 6]

TABLE 6 - AVERAGE WIND DIRECTION AND VELOCITIES (4 years)

NNEESESSWWNW
MONTH% of
Time
Avg. Vel
MPH
%Vel.%Vel.%Vel.%Vel.%Vel.%Vel.%Vel.
JANUARY9.56.512 8.3 9 5.1 3.3 5.9 13 6.0 20 10.1 12 6.8 20 7.7
FEBRUARY116.711 7.9 12 5.8 4.5 6.5 8.3 6.5 17 11.1 16 7.5 21 8.2
MARCH9.87.49 8.3 14 6.0 5.3 8.3 12 7.8 24 14 12 7.9 13 8.2
APRIL116.712 7.6 14 6.1 5.5 5.1 13 7.5 24 10 10 7.3 11 7.7
MAY9.36.56.3 6.5 8 5.2 3.8 6.6 18 6.8 29 8.9 14 7.5 11 6.8
JUNE126.196.38.55.446.6146.5358.5106.275.8
JULY115.6146.2105.34.55.9246.4267.975.925.4
AUGUST145.2125.4154.84.55.1145.1256.77.8554.9
SEPTEMBER146.1267.7165.366.2115.4166.75.35.754.9
OCTOBER236.6186.99.84.735.68.85.916795.1135.9
NOVEMBER176.2137.3125.45.56.2136.7128.4106.2166.9
DECEMBER166.4117.69.85.42.85.67.35.9168.8196.9187.1


[ Table 7]

TABLE 7 - AVG. MONTHLY FREQ. OF WIND DIR. AND AVG. VELOCITIES OVER 7 YEARS
TOTAL NUMBER OF PRECIPITATIONS IN 7 YEARS RELATIVE THERETO At 3AM, 10AM and 3 PM

NNEESESSWWNW
Range of
Wind Velocity
in MPH
Time of
Observance
Avg. Monthly
Freq. of Events
Total Occurrence
of ppt in 7 years
Avg. Monthly
Freq. of Events
Total Occurrence
of ppt in 7 years
Avg. Monthly
Freq. of Events
Total Occurrence
of ppt in 7 years
Avg. Monthly
Freq. of Events
Total Occurrence
of ppt in 7 years
Avg. Monthly
Freq. of Events
Total Occurrence
of ppt in 7 years
Avg. Monthly
Freq. of Events
Total Occurrence
of ppt in 7 years
Avg. Monthly
Freq. of Events
Total Occurrence
of ppt in 7 years
Avg. Monthly
Freq. of Events
Total Occurrence
of ppt in 7 years
03AM.251.081.08.08.25.25.08.08
to10AM.081
13PM.08
23AM8.8134.879.1132.211.8126.847.9127.32
to10AM8.0113.566.611.3323.743.924.162.32
43PM5.643.276.591.322.943.154.653.64
53AM22.8316.56411.0255.31216.824361.619.21620.310
to10AM23.32526.15618.7405.21216.32833.62716.81021.515
123PM16.253220.57316.9368.82220.93232.83718.81019.416
133AM1.331.67.40-.42-0.9035.17.58-1.31
to10AM0.421.24.081.2510.80110.191.212.4-
223PM0.665.832--.5041.379.6151.7-3.62
233AM.08
to10AM.08.08
313PM.25


[ Table 8]

TABLE 8 - TOTAL NUMBER OF WIND DIRECTION AND TOTAL NUMBER OF PRECIPITATION
EVENTS OBSERVED AT 3 AM, 10 PM and 3 PM, FOR ALL WIND VELOCITIES FOR THE PERIOD
1944 - 1950, INCLUSIVE

NNEESESSWWNW
MonthNo. times
wind obs.
No. times
pptn occur
No. times
wind obs.
No. times
pptn occur
No. times
wind obs.
No. times
pptn occur
No. times
wind obs.
No. times
pptn occur
No. times
wind obs.
No. times
pptn occur
No. times
wind obs.
No. times
pptn occur
No. times
wind obs.
No. times
pptn occur
No. times
wind obs.
No. times
pptn occur
January491780296410132829164189081108
February6395519722523853141042187101347
March647581671163511654144177761157
April651574116612286651517615622879
May821051949142869016171171023784
June78765357714255731017911707512
July791371226263661272019121628222
August886831210461927735160127011471
September1171213923802225465121097463460
October17424113237710198496881610671
November971867286311262891611077361108
December94188225581415660111201599611312


[ Figure 1]

Figure 1. Pictorial Plan Of The Raleigh Research Reactor


[ Figure 2]

Figure 2. Longitudinal Section Through Reactor Building Showing Elevations


[ Figure 3]

Figure 3.


[ Figure 4]

Figure 4. Isometric View of Waste Holding Tanks Piping Diagram of Laboratory Drainage System


[ Figure 5]

Figure 5. Horizontal Section of Reactor Assembly


[ Figure 6]

Figure 6. Sketch of the Complete Reactor Assembly


[ Figure 7]

Figure 7. Photograph of the Assembled Reactor Shield


[ Figure 8]

Figure 8. Photograph of Partially Disassembled Reactor Shield


[ Figure 9]

Figure 9. Photograph of the Two Stationary Blocks of the Reactor Shield


[ Figure 10]

Figure 10. Vertical Cross Section Through Exposure Port Showing Underneath Through Tube


[ Figure 11]

Figure 11. Stainless Steel Reactor. ½ Size


[ Figure 12]

Figure 12. Reactor Safety Envelope


[ Figure 13]

Figure 13. Photograph of a Model of the Reactor Cooling Coils


[ Figure 14]

Figure 14. H2-O2 Disposal System


[ Figure 15]

Figure 15. Refrig. Gas Cooler


[ Figure 16]

Figure 16. Catalyst Chamber


[ Figure 17]

Figure 17. Primary Condenser


[ Figure 18]

Figure 18. Gas Withdrawal System


[ Figure 19]

Figure 19 Plan of Control Room


[ Figure 20A]

Figure 20A. Plan of Neutron Measuring Instrumentation


[ Figure 21]

Figure 21. Plan of Neutron Measuring Instrumentation


[ Figure 22]

Figure 22. Control and Shim Rod Mechanism


[ Figure 23]

Figure 23. Five Second Rod Removal


[ Figure 24]

Figure 24. Wind Direction, Velocity, and Percentage Frequency 1945-1948


[ Figure 25]

Figure 25. Wind Direction, Velocity, and Percentage Frequency 1945-1948


[ Figure 26]

Figure 26. Wind Direction, Velocity, and Percentage Frequency 1945-1948


Notes:

a1Proposal of a Nuclear Reactor at N. C. State College, N. C. State College
Bulletin, July 5, 1949, Clifford Beck. [See image of page 1]

a2Program Administration and Installation Design of the Nuclear Reactor Pro-
ject
at N. C. State College, N. C. State College Bulletin, July 5, 1950. [See images of page 1, 2, and 40]

a3Further notes on Characteristics of N. C. State Research Reactor, N. C.
State College Bulletin, September 10, 1950, Beck, Menius, and Murray. [See image of page 1]

a4Letter of October 12, 1950, from K. S. Pitzer, U. S. Atomic Energy Com-
mission
to President Gordon Gray, University of North Carolina. [See image of page 1]

b4"An Enriched Homogeneous Nuclear Reactor", Los Alamos Scientific
Laboratory
, R.S.I. 22, No. 7, p. 489, July 1951. [See image of page 6]

a5Personal correspondence with Dr. L. D. P. King and his staff. [See image of page 6]

a6Suggestion for special materials and the methods of determining mixing proce-
dures
were obtained from Oak Ridge progress reports on the HRE, and from
Dr. A. Kitzes and associates. [See image of page 7]

a7Established by Professor C. R. Bramer, of the N. C. State College Civil
Engineering Department. [See image of page 7]

c1Considerable help on the design and construction of this system was received
from the Chemical Engineering Department of N. C. State College. [See image of page 14]

d1Design obtained in correspondence from Dr. L. D. P. King of Los Alamos. [See image of page 16]

e1Private correspondence from Dr. L. D. P. King. [See image of page 22]

f1Private correspondence from Dr. L. D. P. King. [See image of page 25]

a8K. Way, Phys. Rev. 70, 115 (1946) [See images of page 29, 30, and 43]

a9"Uses and Limitations of a Low Power Reactor in Scientific Research," by Clifford
Beck
. [See image of page 46]