Atomic Power Station
Radioactive Material Control JACQUES R. LAPOINTE and ROBERT D. BROWN Bettis Atomic Power Division, Westinghouse Electric Corp., Pittsburgh, Pa.
Tm
NATION’S first large scale nuclear power station a t Shippingport, Pa., is located in a populous area and considerable engineering effort was needed to design a system for disposal of its radioactive wastes. The power source, of course, is heat generated by nuclear fission in the reactor core, which is continuously removed by recirculating pressurized water, called reactor coolant. Although this water is of high purity, inhibitors are added to reduce corrosion and all surfaces in contact with the water are constructed of highly corrosionresistant stainless steels. Nevertheless, corrosion to the extent of about 10 mg. per sq. dm. per month does occur. Within a conventional power plant, this would be insignificant. However, in a reactor plant, these minute impurities can become activated in passing through the reactor core and constitute the source of induced radioactivity. Also, despite rigid inspections and controlled techniques in manufacturing, a small percentage of fuel elements in the core might possibly develop small holes and release fission products into the coolant. Factors such as maintenance of equipment, normal shutdown, and safety valve discharges require that contaminated coolant be drained and disposed of at the plant site.
wastes will contain no appreciable fertile material and the disposal process need not include provisions for reclaiming valuable constituents. Also, because of the nature of the wastes, hazards and associated problems also are considerably less than those at a radiochemical processing plant. In determining which disposal method would be suitable, several factors had to be evaluated: Has the process been commercially proven or developed at least to pilot plant scale? Is the over-all activity level reduced sufficiently for safe disposal to the environment? Is the process economical, and are operating personnel and people in surrounding areas properly protected?
I.
Table
Predicted Gross Factors
Gross Decay Factor Nonvolatile PWR Wastes
0
4
6
5
No fuel reprocessing facilities are planned for this site; hence, radioactive INDUSTRIAL AND ENGINEERING CHEMISTRY
1.2 2.2 3.2 10 31 73 99
1-3 1-11 1-16 1-3 1 1-46 1-61 1-76
The over-all reduction in the design activity level required for nonvolatile
There are four main methods of radioactive waste disposal
Volatile PWR Wastes E
b Dilution and dispersion
1-3 1-11 1-16 1-3 1
b
Land or sea burial
1-60 1-76
b
Concentration and storage
b
Decay
Decay Interval, Days
Liquid Waste Disposal
Natural decay
Selection of a Disposal Process
980
fission products in the pressurized water reactor (PWR) wastes before they can be discharged to the environment is approximately 2 X 108. This can not be practically achieved by natural decay processes alone. Based on the long-lived radioisotopes, 33-year cesium-137 and 20-year strontium-90, the average holdup time required was estimated at about 80 years (Table I).
1-45
a d
d
1.3 3.7 7.1 50 360 2500 16,700
1 to 24 hours. Shutdown t o 1 hour. Shutdown t o 24 hours. Not known.
Provided an adequate supply of uncontaminated water is available, proper blending with copious amounts of uncontaminated water is probably the least expensive and requires minimum handling. Although the PWR plant is situated advantageously on a fairly large river, it was decided early not to use the Ohio River proper for diluting liquid wastes but to use river water only for cooling the main condenser. This decision was made because of the impossibility of obtaining complete mixing, with the consequent chance of exceeding waste disposal specifications. By properly controlling the discharge of low-level wastes, the limited dilution process utilized is entkely under the close supervision of the station operator who is provided with suitable alarms and instruments to protect against accidental discharge. Furthermore, the river, which provides some further dilution, serves as an additional design safety factor. Disposal by Ocean Burial
Disposal of all radioactive wastes by ocean burial (5) was explored. The procedure, similar to that used by other Atomic Energy Commission installations and modified to suit these special conditions, involves the mixing of waste liquor as received from the reactor plant with dry cement in a 55-gallon steel drum and allowing the mixture to solidify. The drums would be trucked to a marine terminal, picked up by a vessel, and buried at sea. Because of the predicted large quantities of radioactive liquid wastes, this method without concentration was too expensive-$3.50 to $10 per gallon with as much as 30,000 gallons of waste to be drummed each month. Radioactive waste liquors can also be concentrated by evaporation with subsequent land or ocean burial of the residual liquor. This is done a t various AEC installations. Practical designs have been developed for evaporators where activity of the condensate is less than one millionth of the activity of bottoms liquor (decontamination factor, 106). Essentially, this has been done by highly efficient deepbed vapor filters to remove entrained liquid particles. Close pH adjustment of wastes (slightly alkaline) adequately prevents iodine and ruthenium in the feed solution from volatilizing. In spite of the merits of evaporation, it is relatively costly (10 to 13 cents per gallon) and complicated by remote disassembly techniques. For these reasons, evaporation of all liquid wastes was ruled out. However, a limited vapor compression evaporation process appeared feasible for processing high chemical content wastes. This liquid cannot be processed through the ion
>ZOO m r l h , CONTACT DOSE OCEAN B U R I A L Y INCLUOES E V I P O R I T O R OVERHEAD PRODUCT
Radioactive waste disposal facilities for handling reactor plant effluent at the Shippingport Atomic Power Station
exchange system without sacrificing resin or resin bed performance. Ion Exchange
Decontaminating liquid waste by passing it through ion exchange resins has been the basis for much AECsponsored research ( 3 ) . The effectiveness of ion exchange resins for removing fission product radioisotopes from various laundry and chemical process wastes has been established. However, experimental data were lacking for the decontamination factors that could be achieved, and the number of column volumes of waste that could be processed before activity break-through would occur, with the wastes consisting of pure, deionized water containing carrier-free radioisotopes. The reactor coolant discharged from a heterogeneous pressurized water reactor, such as the PWR, is a source of such pure waste because it is continuously purified to keep its electrical conductivity a t 1 micromho per cm. or better during operation. Because of the merits of ion exchange, an evaluation research program was instituted a t the Bettis plant. Results, based on in-pile tests of defective PWR type fuel elements, demonstrated that decontamination factors obtainable when processing PWR-type wastes are a t least as great as those required by the PWR waste disposal system design. Also, large volumes of these wastes could be processed by ion exchange before activity break-through with negligible resin consumption. As a result of this evaluation research
program and earlier studies made by the Bettis plant, ion exchange was chosen as the principal method of high purity liquid-waste disposal.
Solid Waste Disposal Spent Demineralizer Resin
One method considered was flushing the spent resin out of the reactor plant demineralizers into a filter, and then conveying in bags or containers to a concrete vault. This disposal method was discarded because it would require remote handling techniques and expensive handling equipment. Also, a watertight steel liner must be provided for the concrete vault to prevent seepage into the ground as well as a means for removing decay heat from the solid material. Resin incinerating was also considered, but the need for intermediate storage was disadvantageous. Other apparent disadvantages : heavy shielding for the incinerator, and special washing and treatment of the radioactive gases before discharge to the atmosphere. For these reasons, this method was not adopted. Resin flushing from demineralizers directly into an underground storage tank was another method considered. Remote handling or other equipment would not be needed. The resin would be allowed to settle and the surface water above the resin-water interface removed for further processing. This method of resin waste disposal was selected because it is the simplest, safest, and most economical in capital and operating costs. VOL. 5 0 , NO. 7
JULY 1958
981
'
I
1
REACTOR P L A N T SPENT RESIN 36 cu f l / m o n l h
a
i o 4 w w C c overage
1
I
-
pq NON-COMBUSTIBLE
COMBUSTIBLE WASTES
Gaseous Wasfe Disposal
I
2 0 0 0 - 4 0 0 0 I b f m o av9
ITEMS
TOO L A R G E
Dilution and Dispersion
TO S H I P
L
\ / /
OCEAN BURIAL SITE BURIAL
SPENT R E S I N S T O R A G E AT S I T E
Radioactive waste gases are generallv disposed of by filtering, diluting with air, and discharging to the atmosphere fi om a high stack. This is standard procedure at AEC installations where fissionproduct gases are released during processing of spent fuel elements Disposal of waste gases from the PWR plant by air dilution alone would require a volume of air so large it would be impractical. Liquification is another disposal method considered practical but it \\as still in the pilot plant stage at the time the PWR s)lstemwas designed. ,4 practical solution for the disposal of these waste gases, since they are principally 5.3-day xenon-133 and, to a lesser extent, 10-year krypton-8.5, is decay and dilution. This reduction, coupled with air dilution, would reduce the gases to the allowable tolerance (Table I).
This i s whaf happens to radioactive solid wastes at Shippingport
Combustible Waste
A study was made of combusrible waste disposal by baling dry waste in a standard baling machine, and placing in a 55-gallon drum which would then be filled with concrete and dumped into the ocean. The estimated cost for this method was 4.5 cents per pound of waste.
The incineration method was also studied because it has the added advantage that damp waste and wood can be handled satisfactorily; this is not always true for baling. Although the capital cost of equipment would be much larger than for baling, the cost per pound of waste was estimated at 30 cents. Therefore, this method was selected.
Cost of the Atomic Power Station Is Estimated at $72,500,000 $1,000,000
Heat source Core Reactor vessel Reactivity control and handling Subtotal
11.9 2.1
core
Plant Reactor coolant system Auxiliary systems Instrumentation and control systems Special test equipment, spares, manuals Installation Subtotal Structures and facilities Plant container Secondary shield and fuel canal concrete Fuel handling and service buildings Internal shielding and plant container drainage Subtotal Architect-engineer Expanded hot lab. facilities Contingency Subtotal nuclear portion Estimated turbogenerator plant Total
--3.0
17.0
3.9 5.8 1.9 0.5 13.2 -___ 25.3
2.4 4.9 1.9 0.8 -
10.0 1.8 0.5 __0.4 55.0 17.5 __ 72.5
Design Disposal Criteria
The vapor compression evaporator unit, located in a shielded vault beside the waste disposal building, has a capacity of 100 gallons per hour
982
INDUSTRIAL AND ENGINEERING CHEMISTRY
Radioactive waste disposal problems are similar to those for chemical or sanitary wastes. Maximum permissible concentrations (MPC) have been established (4, 6) and a major aspect of design is to ensure that these values are not exceeded (Table 11).
RADIOACTIVE MATERIALS Specific PWR Criteria
Radioactive waste disposal facilities for PWR were designed to meee stringent engineering limits on the activity levels leaving the site ( 2 ) . Major criteria were established--e.g., wastes must be processed, when necessary, so that their concentration in water, measured a t the condenser water stream before discharge to the Ohio River, and in air, measured a t the point of discharge to the surrounding atmosphere, shall not exceed 'one tenth of the maximum permissible concentrations recommended in Table 11.
Quantity and activity were unknown for all radioisotopes in the PWR coolant during the design stages; therefore, the design for disposal of nonvolatile material was based on. one tenth of the more restrictive value listed for an unidentified mixture of radioisotopes rather. than the more liberal values for specific isotopes (Table 11). Also, released material must conform to requirements given for disposal of radioactive wastes in Codes of the Commonwealth of Pennsylvania (7). Tank or burial facilities must be sealed against leakage into the ground and vapor release into the atmosphere, and sufficient capacity must be provided to store radioactive gases for the maximum duration of a weather inversion. This period, based on the Donora incident (8)is 5 days for the area. Wastes shipped must be packaged in a manner to prevent a radioactive hazard to packing and shipping personnel (9).
A THESE BARRIERS PROTECT PERSONNEL MATERIAL
b Fuel element cladding-corrosion-resistant
FROM RADIOACTIVE
zirconium alloy
b Aq, all-welded pressurized system for cooling the elements. fuel elements can b e located and removed if coolant activity should become too high
Faulty radio-
b The reactor plant container, a group of interconnected vapor-tight steel pressure vessels, which houses the entire reactor plant of the pressurized water reactor
Description of Waste Removal Process Radioactive wastes treated at the PWR plant are classified as solid, liquid, and gaseous. Solids are subdivided into combustible and noncombustible wastes; liquids are subdivided into reactor plant effluents, monitored wastes, special monitored wastes, and decontamination-room wastes. Each classification and subdivision requires a somewhat different treatment, but several basic processes are used-Le., natural decay, removal by ion exchange, separation by evaporation, and retention Table 11. Maximum Permissible Concentrations of Typical Radioisotopes" ro./Ml. Radioisotope 33-yr. Cs137 19.9-yr. Sr9O 54-day Sr89 67-hr. Moo9 8-day 1181 5.3-day Xe133
(4, 6,7 ) .
Water 1.5 x 10-8 s x 10-7 7 x 10-6 14 3 x 10-6 4 x 10-8
Air
x 10-7 2 x 10-10 2 x 10-8 2 x 105 x 10-0 4 x 10-6 2
in underground storage tanks at the plant site. Some wastes require only one of these processes but others require a combination. Reactor Plant Effluents
Reactor plant effluents may contain highly radioactive soluble and insoluble materials as well as dissolved radioactive gases. The estimated average volume of these waters is 3156 cubic feet per month. Reactor plant effluents are treated by decay, ion exchange, and separation and decay of the soluble gases from the liquid. The gross activity of these wastes comes from both volatile and nonvolatile materials. The gaseous or volatile activity is estimated a t a maximum of 8.7 pc. per ml. of liquid, whereas the nonvolatile activity is estimated a t a maximum of 1.4 pc. per ml. of liquid. After treatment, the radioactivity is estimated a t a maximum of 1.2 x 10-8 for volatiles and 4,5 X 10-6pc. per ml. for nonvolatiles. The waste is monitored in the
system's test tanks and the proper flow rate determined so that after blending with the turbine condenser cooling water stream the activity will be no more than 1 X 10-8 pc. per ml. This is one tenth of the maximum permissible concentration for an unidentified mixture of isotopes. Liquid waste is sampled and the radioactivity determined a t each stage of a treatment that includes a series of batch processes. The waste may be reprocessed if the samples indicate inadequate radioactivity reduction. Reactor plant effluents are piped to surge and decay tanks where they are stored for a 45-day period. During this interval, radioactivity content will decrease by a factor of 31 for the nonvolatiles and 360 for the volatiles. The tanks, of Type 347 stainless steel, are located underground to give protection against radiation. The decayed liquid is cooled and then circulated through a series of four ion exchangers where nonvolatile radioactivity will be reduced by a factor of 1000. The radioactive gases VOL. 50, NO. 7
JULY 1 9 5 8
983
INSTRUMENTATION AT THE SHIPPINGPORT POWER STATION directly to the reactor coolant. The indicating instruments for these quantities are grouped on the main control console of the reactor plant. Signals for control and alarm circuits are also supplied b y primary plant instrumentation. Thermocouples sense bearing and operating temperatures of pumps and other plant components. Primary loop temperatures are sensed for basic control purposes b y five resistance thermometers in each o f the four coolant loops. Four of these are fast response thermometers for plant control and information. The fifth thermometer in each loop i s a highly accurate unit that calibrates the fast response units. Each coolant loop also contains instruments for measuring flow, using the pressure drop across a calibrated venturi, and for measuring pressure b y differential and static pressure units. Any portion of the reactor coolant system that may be isolated b y valves has a full-range pressure instrument for operation information and, in some cases, for control. Units on the reactor vessel provide automatic shutdown on low pressure.
The Reactor The reactor has two instrumentation systems-nuclear instrumentation which reports the nuclear reaction, largely for control purposes; and core instrumentation which i s for information only. Nuclear instrumentation is the most important. It monitors the neutron flux level of the core, which when averaged in neutrons per square centimeter per second measures directly heat output o f the core. Measurement of neutron flux values is a representative sample, not an absolute measure. Moreover, the neutron detectors must be located outside the pressure vessel because of the neutron flux level range (ratio, approximately 1 O9 to 1) and because of the formidable problem of penetrating the pressure vessel. For accurately monitoring the neutron flux level over its entire range, two types of detectors are used: a boron trifluoride (BFa) proportional counter that reports from below source level to about 10,000 times source level; and a compensated ionization chamber which includes the range from 1000 times source level to 1 billion times source level.
Remote V i e w i n g System Radioactivity Monitoring Radiation i s monitored b y two separate systems, one for measuring power level, the other for safety. The power level monitoring system provides information necessary for operating the reactor plant within the radiation levels required b y the Atomic Energy Commission. It has several channels, each containing one or more detectors, preamplifiers where necessary, and a computer indicator providing read-out. The safety radiation-monitoring system determines any increases in the normal background of the plant surroundings. All equipment necessary to locate and measure radia-
Primary Plant Instrumentation System This system includes all conventional instrumentation associated with the main and auxiliary fluid systems of the primary, or reactor, plant. Measurement of primary plant parameters such as flow, level, pressure, temperature, oxygen concentration, and acidity is necessary for safe and reliable operation, and for proper maintenance of the primary plant. The most important of this instrumentation is concerned with quantities relating
are then removed in a gas stripper and the volatile activity is reduced by a factor of 2 X 106. Liquid waste from the stripper is then piped to one of two 5000-gallon carbonsteel test tanks which are plastic-lined and located a t ground level. Shielding is not required since the radioactivity has been reduced to a very low level. The liquid is held in these tanks until analyses disclose proper amount of radioactivity reduction. During normal operation one test tank is filled while the
984
tion hazards are included in this system. Areas normally occupied within the plant buildings will be monitored to warn operating personnel against increases in radiation levels. Hand and foot monitors will also be used, and monitors will b e provided at exits. Film badges and dosimeters will be provided for all personnel to record cumulative doses of radiation.
other is emptied to the turbine condenser discharge stream through a flow regulator and radioactivity monitor. The flow rate and radioactivity are continuously recorded. The waste liquid is then discharged into the condenser cooling water through a distributor located in the plant effluent channel. The blended liquids pass over a weir for further blending. A sampling rake is located downstream from this weir. Here samples are taken from 21 different locations in the
INDUSTRIAL AND ENGINEERING CHEMISTRY
This system i s composed of two parts. One part has five television cameras to view areas within the chambers during operat:on. One is located in each of four boiler compartments and one in the pressurizer compartment. Camera tilt, swing, and focus are remotely adjustable from the control room. One monitoring receiver is used; i t can be switched to any one of the five cameras. The other part of the system has four fixed-position, fixed-focus cameras adjusted to view the four boiler drum level gage glasses, and four television monitors, one for each camera in the control room.
channel. These are mixed and analyzed to ensure that the radioactivity of the liquid leaving the plant meets the criteria established for the plant’s disposal system. Monitored Wastes
Waste from showers and cold (nonradioactive) laundry under normal conditions is not expected to be radioactive; nevertheless, it is monitored in tanks as a safety precaution. The estimated volume of this waste is 22,600
R A D I O A C T I V E MATERIALS cubic feet per month and the estimated maximum radioactivity (all nonvolatile) is 8.1 X 10-7 pc. per ml. After monitoring, the waste is discharged to the effluent channel where it is diluted with the turbine condenser cooling water and discharged into the Ohio River.
Wastes from the “hot” laundry and laboratory drains are pumped into tanks in the waste disposal area. The estimated volume is 11,850 cubic feet per month and the estimated maximum activity (nonvolatile) is 2.16 X l o 4 pc. per ml. After monitoring, the waste is pumped to the coolant channel a t a controlled flow rate to prdduce a blended radioactivity of 10-8 pc. per ml. or less. If the radioactivity is too high for the tank to be emptied in normal time, the waste may be transferred to the decontamination room waste tanks for evaporation.
gas surge tank. At infrequent intervals the net gas made from the system is discharged to one of four gas-decay drums and stored a t 50 p.s.i.g. long enough (60 days) for radioactivity to decrease sufficiently for controlled discharge and dilution with 9000 cubic feet per minute of air. Gross activity of the gaseous wastes leaving the stack does not exceed 4.0 X 10- pc. per ml. of air which is one tenth of the maximum permissible concentrations for a mixture of these isotopes of xenon and krypton. All gases from a breathing system for the resin storage and surge and decay tanks are discharged into the vent gas system which not only provides storage but also maintains constant pressure in the vapor spaces of these tanks. The major equipment in the gaseous waste system includes the vent gas surge drum, four gas decay drums, three gas compressors, the gas stripper, and two airdilution fans.
Decontamination Room Wastes
Combustible Solid Wastes
Effluents from the decontamination or cleaning rooms are pumped to tanks in the waste disposal building. The estimated volume of this waste is 1025 cubic feet per month and the estimated maximum activity is 0.27 pc. per ml. This stream may contain a high level of dissolved solids which will be concentrated, mixed with cement in standard 55-gallon drums, and then buried a t sea. The decontamination room waste stream is first neutralized with caustic soda or sodium carbonate and then piped into a commercial vapor compression-type evaporator. The evaporated liquid (overhead) although essentially free of radioactivity is piped to the surge and decay tanks where it is processed as part of the reactor plant effluents. This safety precaution precludes accidental dumping of high activity liquid into the effluent channel. The concentrated liquid accumulates in the evaporator until there is enough to fill a drum. The stainless steel evaporator has a capacity of 100 gallons per hour and is located in a shielded vault beside the waste disposal building.
Combustible solid wastes are burned in an incinerator. The flue gases are sent through a wet gas scrubber which removes particulate matter 5 microns and larger and also cools the gas. The cooled flue gas is then passed through an exhaust filter for final cleanup before discharge to the stack. The incinerator residue ash is dumped into the ash slurry tank where it is mixed with water and pumped to the resin storage tank. An estimated 2000 to 4000 pounds of combustible waste will be processed each month, but no realistic estimate can be made of its radioactivity.
Special Monitored Wastes
Gaseous Wastes
Most of the radioactive gases, predominantly 5.3-day xenon and 10-year krypton, formed in the reactor remain dissolved in the effluent during the waste treatment process. In the gas stripper which operates under a slight vacuum, the gas is stripped from the solution and sent to the vent gas system. An insignificant amount of gas remains in the solution. The gas from the top of the stripper is compressed and then piped to a large
Noncombustible Solid Wastes
The noncombustible solid wastes are resins from the plant’s ion exchangers, residue ash from the incinerator, solids from strainers in the pipe lines, and contaminated plant equipment. Equipment is encased in concrete and buriedsmall items a t sea, and large items a t the site. Resins, residue ash, and strainer solids are mixed with water and pumped to one of two resin storage tanks. When solids have settled, the surface water is pumped to one of the surge and decay tanks and then processed in the same manner and in the same equipment as the reactor plant effluents. Solid, noncombustible material that will be placed in the resin storage tanks is estimated a t a maximum of 36 cubic feet per month. These stainless steel resin storage tanks are located in individual, totally enclosed, underground vaults several feet under ground. The incinerator and ash slurry tank, of stainless steel and located in a pit in the waste disposal building, are operated remotely.
Control of the Process System
Waste streams will be sampled as they enter the waste disposal system, before and after each treatment, and before discharge from the plant. Liquid and gaseous diluted waste streams can be monitored, and any waste stream may be reprocessed if necessary. This flexibility, along with the basic batch-type of processing and the large number of sample locations will minimize the possibility of discharging wastes with more than the limit of radioactivity. Coolant activity monitors, installed on the main liquid coolant header and on the stack, detect and inform of accidental discharge of either liquid or gaseous wastes above specified limits.
.
Off Site An off-site monitoring program was started early in 1956 and will continue during operation of the plant. Types and amounts of radioactive materials that occur in the environment around the PWR plant site are determined; this was done for about 11/2 years prior to start-up of the nuclear portion of the plant. Thus, when the plant is in operation, significant changes resulting from plant operations can be readily detected. Soil in the general vicinity of the plant, Ohio River water a t two locations above and one below the site, well water within a 1-mile radius, and both vegetation and air in this general area are being tested. Also, another monitoring station for river water has been installed a t the “first point of use” below the plant, which is the waterintake service to the city of Midland, Pa. Radioactive dusts and gases in the air, and normal background radiation in the vicinity are monitored by five stations-four fixed and one mobilewhich are located both downwind and upwind of the plant site. The mobile station will determine the relationship between the four fixed stations and other areas of interest. Mud samples, taken from the Ohio River bottom and from pools behind dams near the plant site, together with river algae are compared with similar samples when radioactive wastes are being discharged. Acknowledgment
The authors wish to express their appreciation to the U. S. Atomic Energy Commission, the Duquesne Light Co., and the Westinghouse Electric Corp., for permission to publish the information contained in this paper. They also wish to express their gratitude to J. V. A. Longcor and W. R. Kennedy of the Stone and Webster Engineering Corp. for their assistance in designing the PWR radioactive waste disposal plant. VOL. 50, NO. 7
JULY 1958
985
Safeguards at Shippingport All conceivable accidents that could threaten the safety o f the pressurized water reactor plant have been analyzed-e.g., loss of station power; loss o f coclant flow to the core; reactivity accidents such as uncontrolled rod withdrawal or cold water entering the core and causing increased reactivity; loss of reactor coolant if a rupture occurred in the primary coolant system; and accidental release from the radioactive waste disposal system. Conclusions reached from these studies are: Even a complete loss of electrical power would create no hazard, either to personnel within the plant or to people in the surrounding area. The reactor protection system i s designed to prevent any damage to the core; no combination of pump failures or other loss o f coolant flow can release fission products from the fuel elements. The system precludes any nuclear transient that would result in vaporizing or gross melting of the fuel elements with an attendant large pressure build-up in the reactor coolant system. Consequently, a nuclear excursion could not both rupture the primary coolant system and simultaneously release significant quantities of fission products from the fuel elements. This con-
References (1) Hatch, L. P., others, Nucleonics 12, 14
(1954).
(2) LaPointe, J. R., WAPD-T-419, Uni-
versity of Michigan, Ann Arbor, Mich. (3) Lindsay, W. T., Abrams, C. S., WAPD-CP-2636.
(4) National Bureau of Standards, Handbook 52, U. s. Government Printing Office, Washington, D. C . , 1953. (5) Ibid.,Handbook 58,1954. (6) Ibid.,Handbook 61, 1955. (7) Pennsylvania Department of Health, Radiation Protection Regulation 433, Harrisburg, Pa. (8) Schrenk, H. H., others, U. S. Public Health Bull. 306, Federal Security Agency, Washington, D. C., 1949. (9) U. S. Interstate Commerce Commission, Codeof Federal Regulations, Title 49, Parts 71 to 90, U. S. Government Printing Office, Washington, D. C., 1956. RECEIVED for review July 25, 1957 ACCEPTEDMarch 20, 1958 Work done under prime contract of the U. S. Atomic Energy Commission to further the peacetime use of atomic energy in electric power generation.
986
INDUSTRIAL A N D ENGINEERING CHEMISTRY
clusion takes into account the possibility of a chemical reaction of zirconium with water. Release of 100% of the yeactor coolant to the outside atmosphere would not result in a biological hazard at the site boundary. This would be true even if the core had been operated for 3000 hours a t full power and the coolant contained the maximum activity that could be caused b y imperfections in the cladding of about 1% of the uranium dioxide fuel elements. About 1000 leaky elements are expected during the entire life of the core. Complete loss o f reactor coolant through a major rupture in the reactor coolant system i s the only accident which could release a significant amount of fission products to the plant container. However, should this improbable accident occur, the safety injection system can pump enough water into the reactor coolant system to prevent core melting and subsequent development of a biological hazard beyond the site boundary. This applies to all locations and sizes of rupture, except a rupture larger than 6 inches in diameter in the reactor coolant system pressure boundary below the level of the core and in the immediate vicinity of the reactor vessel. In that event, some core melt-down, zirconiumwater reaction, and release of fission products to the plant confainer could occur. Based on the minimum expected flow of 1500 gallons per minute from the safety injection system, a maximum of 2OY0 of the blanket region of the core could melt. However, the biologically significant fission products which would escape from the core to the container would represent only an estimated 1.4% of the total activity in curies at shutdown. The major portion of the fission products would probably remain within the fuel material during melt-down and only a very limited quantity of fission products would be released. During such an accident, a maximum of 4.5% of the 9.38 tons of zirconium associated with the heat transfer areas in the core could react with the steam or water. The zirconium-water reaction could not start for some minutes after the rupture and would be in the nature of a self-propagating fire.
The pressurized water reactor plant container can adequately withstand the maximum pressure that could develop from the most extreme rupture in the coolant system-i.e., one releasing all reactor coolant and secondary water from one boiler. Heat from any subsequent exothermic zirconiumwater reaction and from combustion of the hydrogen produced by this reaction would only slightly increase pressure within the plant container (pressure vessels). Total pressure would still be considerably less than the container was designed to withstand. If the hydrogen gas does not burn as it is evolved, but mixes uniformly with the air and steam within the plant container, the hydrogen concentration will be below the flammable limit and neither burning nor detonation will occur. The total external gamma radiation dose accumulated during a 12-hour period as a result of seepage from the plant container at a point on the site boundary 1700 feet distant is calculated to be about 1.0 roentgen. Anyone directly in the path of leakage at 1700 feet for 12 hours would receive an integrated dose to the thyroid gland, through inhaling the iodine fission products, of about 500 roentgen equivalent physical units (rep). The integrated 3-year dose to the bone from inhaling strontium fission products would be about 0.20 rep. In “Theoretical Possibilities and Consequences of Major Accidents in Large Nuclear Power Plants,” March 1957, the Atomic Energy Commission has defined an acceptable emergency dose as one from which persons are assumed to have received no injury and is taken to be 25 roentgens of wholebody gamma radiation in one brief exposure or 50 roentgens in 3 months. This reference also sets 2000 rep to the thyroid gland and a lifetime dose to the bones of 50 rep as reasonably acceptable e m s gency doses. Therefore, even the extreme loss-of-coolant accident presents no biological hazard to the uncontrolled area surrounding the site. No significant radiation hazard would result from the maximum possible accidental release of gaseous or liquid wastes from the radioactive waste disposal system.
