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

%NEfurther010052; %NEfurther010052forms; ]> Further Design Features of the Nuclear Reactor at North Carolina State College Machine readable transcription Beck, Clifford; Menius, A. C., Jr.; Murray, R. L.; Underwood, Newton; Waltner, A. W.; Webb, George Creation of machine-readable version: Russell S. Koonts Creation of digital images: Russell S. Koonts Conversion to TEI.2-conformant markup: Russell S. Koonts 187 kilobytes NCSU Libraries Raleigh, NC Further Design Features of the Nuclear Reactor at North Carolina State College

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NCSC #46 Copy 29 of 50, Series A FURTHER DESIGN FEATURES<lb/> OF THE<lb/> NUCLEAR REACTOR<lb/> AT<lb/> 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 Introduction . . . . . . 1 I. The Reactor Building A. General . . . . . . 2 B. Ventilation and Heating System for the Laboratories . . . 3 C. Sewerage System for the Laboratories . . . . . . 4 II. Reactor Design A. General . . . . . . 6 B. Concrete Shielding . . . . . . 6 C. Fissionable Fuel . . . . . . 8 D. The Fuel Cylinder and Reactor Envelope . . . . . . 9 E. Reflector, Thermal Column, Lead Shielding . . . . . . 10 F. Cooling System . . . . . . 11 G. Gas Disposal Unit . . . . . . 13 1. Hydrogen Oxygen Recombination System . . . . . . 13 2. Disposal to the Atmosphere . . . . . . 15 H. Instrumentation and Control . . . . . . 15 1. General . . . . . . 15 2. Fission Chambers . . . . . . 16 3. Neutron Flux Measurements on Linear Scales . . . . . . 16 4. Neutron Flux Measurements on Logarithmic Scales . . . . . . 18 5. Rate of Change of Neutron Flux: Period Measurements . . . . . . 18 6. Gamma Compensated Chamber . . . . . . 19 7. Measurement of the Gamma Ray Flux from the Reactor 19 8. Control Console and Recorder Rack . . . . . . 19 9. Control Safety and Shim Rods . . . . . . 21 10. Campus Monitoring . . . . . . 23 11. Miscellaneous Monitors . . . . . . 23 III. Reactor Characteristics A. Safety Features . . . . . . 25 B. Reactivity and Nuclear Behavior . . . . . . 26 C. Power Level, Radiation Fluxes . . . . . . 28 D. Fission Products; Radioactivity . . . . . . 29 IV. Reactor Hazards and Safety Preacautions A Normal Hazards . . . . . . 36 B. Minor Incidents . . . . . . 38 C. Major Catastrophe . . . . . . 41 V. Operation and Experimental Program A. Start-Up . . . . . . 45 B. Initial Experiments . . . . . . . 45 C. Subsequent Experiments . . . . . . 46 D. Procedures and Policies . . . . . . 46

LIST OF FIGURES AND TABLES

Figure 1 Plan of Reactor Building 2 Elevation Plan of Reactor Building 3 External Appearance of Reactor Building 4 Laboratory Drainage System 5 Horizontal Cross-Section of Reactor Assembly 6 External View of Reactor Assembly 7 Photograph of Assembled Reactor Shield 8 Photograph of Partially Disassembled Reactor Shield 9 Photograph of Completely Disassembled Shield 10 Section Through Typical Exposure Port 11 The Reactor Cylinder 12 The Reactor Safety Envelope 13 Photograph of a Model of the Cooling Coils 14 Hydrogen-Oxygen Recombination-Recirculation System 15 Refrigerated Gas Cooler 16 Catalyst Chamber 17 Primary Condenser 18 Plan of the Gas Disposal System 19 Plan of the Control Room 20 Plan of Neutron Measuring Instrumentation: Linear Systems 21 Plan of Neutron Measuring Instrumentation: Logarithmic Systems 22 Control and Shim Rod Assembly 23 Reactor Behavior as the Control Rod is Withdrawn

Table 1 Characteristics of the Reactor Fuel 9 2 Data Relating to Fuel Cylinder 9 3 Data on Graphite Used in Reflector and Thermal Column 10 4 Activity Induced in the Reactor Cooling Water 12 5 Calculated Radiation Fluxes at 10 Kw Operation 28

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. The Atomic Energy Commission has expressed approval of this project 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.

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" (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.

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.

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

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.

II. REACTOR DESIGN A. General.

Since 1944, the Los Alamos Scientific Laboratory has operated a low-power nuclear reactor of the "water-boiler" type. 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.

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

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. 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 procedures 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

"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

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

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

Volume of Reflector - 105 cu. ft. Weight of graphite in Reflector - 5.4 tons Volume of thermal column - 110 cu. ft. Weight of graphite in thermal column - 5.7 tons Density of graphite - 1.65 g/cc

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.

TABLE 4 - ACTIVITY INDUCED IN THE REACTOR COOLING WATER

TargetIsotope Form ofthe Im-purity Avg.Conc.ppm ActiveIsotope HalfLife λsec-1 σabarns Abund. ofTargetIsotope(%) Ncm-3 Result.Activityd/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

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

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 Reactor 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:

1. The volume of the system and the rate of circulation are such that the hydrogen and oxygen content are maintained below the explosive limit (4% H2) at all times and at all points, even within the reactor. 2. A refrigerated condenser immediately above the reactor (gas tempera- tures lowered to 13oC) returns all vapors to the reactor as condensate. Many fission products such as I2 and Br2 which otherwise would escape, are kept in the reactor. This condenser also prevents liquid contamination of the remain- ing parts of the system, which is particularly important to the "longevity" of the catalyst. 3. Recirculation of the gases through the catalyst provides increased op- portunity for recombination of hydrogen and oxygen, and also provides a consider- able hold-up of radioactive inert gases, so that short-lived activities are con- siderably decayed before discharge.

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.

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

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. 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

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.

Other features of these three channels are: (a) A Master Range Changing Switch permits simultaneous change from one range to another on all three of the above channels. (b) Provision is made on the record to indicate the range being used so that there is no ambiguity as to the actual value recorded. (c) The overall response time for these systems is of the order of 15 seconds for full scale response of the galvanometer and of the order of 4 seconds for the recorders. (d) The recorder is a curve-drawing Brown Recorder.

(e) The indicating meters on the reactor shield are G.E. Type DB 18. (f) A time delay and interlock are provided such that the control rods can not be raised until the linear systems are in operating condition.

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

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:

(1) A galvanometer showing neutron flux (3 (a), above). (2) A meter showing the temperature of the Reactor solution. (3) Control and safety rod positioning switches. (4) The alternate manual-to-automatic control switch. (5) A button for manual release of the control and safety rods. (6) A cathode raytube for monitoring automatic control (3 (c), above). (7) Lights to indicate extreme positioning of control and safety rods. (8) Switches for master range-changing mechanism. (9) Emergency evacuation signal.

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.)

(1) Temperature of reactor coolant (measured at 4 different points). (2) Control Rod position. (3) Reactor solution temperature. (4) Neutron flux, linear.

(5) Neutron flux, logarithmic. (6) Period of the reactor. (7) Campus radiation monitor (3 stations). (8) Campus radiation monitor (3 stations). (9) Recombiner temperatures (coolant, catalyst). (10) Spare (11) Spare (12) Spare

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

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 Reactor, 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.

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

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

A Stack Monitor, to measure the level of activity in the gases being discharged to the atmosphere. A Liquid Monitor at the Sump, to measure the activity in liquids collecting from the Reactor Room drains. Two Liquid Monitors in the drainage system of the laboratories. Two Liquid Monitors in the waste line carrying the effluent reactor cooling water. A Hand and Foot monitor for laboratory personnel. Personnel electroscopes and film badges

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 Reactor 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

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

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

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)(γ/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" fromreactor) 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

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, the amount

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, 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

At the end of 10 seconds, 402 watts At the end of 100 seconds, 253 watts At the end of 1 hour, 124 watts

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

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 τ, the solution of the above equation is

For the down-time between τ 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 - τ + 1, t - τ + 2, etc. inserted, a geometric series

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/λ1 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 λ 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.

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