Howard G. Hembree, U. S. Atomic Energy Commission, Washington 25, D. C.
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AEC Safety Research Program — Reactor Design Factors The millions of people who live in the cities of America are not going to be endangered through operation of nuclear reactors. The reactor safety research program of the Atomic Energy Commission helps make this assurance possible
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UGLEAR REACTORS used for the production of power or for any other purpose are so constituted that they cannot, under any circumstances, produce a bomblike explosion. There is a possibility, however, that they can be hazardous in a different manner—through potential release of fission products to the environ ment. A postulated nuclear inci dent can be used to illustrate what the commission is doing about the possible hazard through its reactor safety research program. An atom of uranium-235 will fission, or split in two, when bom barded with neutrons under certain conditions. For each fission approxi mately two additional neutrons are produced, with fission fragments, γ-rays, and heat. The two neutrons are used to produce further fissions and make up the neutron losses in herent in a nuclear reactor. A condi tion of equilibrium can thus be achieved among neutron production, utilization, and loss, so that the chain reaction perpetuates itself at a spe cific rate and with attendant pro duction of heat. The fission fragments are isotopes of a spectrum of elements—mostly radioactive. Herein lies the poten tial problem. Iodine-131, with a halflife of 8 days, is a thyroid seeker; strontium-90, with a half life of 20 years, is a bone seeker. These two radioactive isotopes are particularly troublesome because they can cause harmful biological effects. The primary safety objective, then, in the operation of nuclear reactors is to prevent release of dangerous radioactive fission products.
The Postulated Accident
Categories of Research
To achieve a safe and conserva tive basis of design for nuclear reac tors, the designers postulate, in the language of the nuclear safety engi neer, a "maximum credible acci dent." The reactor is then built to withstand this accident, and thus ensure the public health and safety. An example of such a postulated accident is this: Assume that the control rods of a solid fueled hetero geneous reactor are by some im probable means withdrawn too far, the reactor becomes too supercritical, and a nuclear excursion or runaway results. Because the circulating cool ing water is unable to remove heat fast enough, the nuclear excursion could melt the uranium fuel ele ments and their aluminum cladding. It is likely that molten metal would contact the water coolant and result in an exothermic reaction.
The probability of an accident such as postulated is exceedingly small. But reactor builders pur posely assume a serious but credible accident, to establish a conservative basis for reactor design. They then design safety features into the reactor to nullify the effects of the postu lated accident. This research falls into three logi cal categories :
In the unlikely event of our postu lated maximum credible accident, the energy that could be released in the nuclear excursion plus that from the exothermic chemical reactions could cause the reactor pressure ves sel to rupture and lead to the escape of fission products known to be present in the reactor core vessel. Assume that the reactor pressure vessel was indeed ruptured by the nuclear and chemical energy releases but that the fission products were safely retained in the vapor con tainer. Unfortunately, there is yet another hazard possible—the hy drogen evolved in the metal-water reaction could explode, rupture the vapor container, and release the fission products to the atmosphere. I/EC
1. Reactor Kinetics and Con trol. Concerned with the kinetic behavior of reactors. Investigates the inherent nuclear character istics that might permit a reactor to run away, and equally impor tant, the inherent characteristics that would shut it down. 2. Potential Reactions. Reac tions between molten metals and water, metals and gases, and, of course, fission product release from irradiated fuel. 3. Containment. Integrity of the reactor core vessel and the outer vapor containment shell. The dynamic loading that would be applied to both the pressure vessel and the containment sphere in the event of an accident must be determined. Means must be devised to ensure leak-tightness of the containment shells. This re quires considerable research, be cause the requirement for vessel integrity is greater in the nuclear business than in other industries. This research program does not include the safety aspects of other commission programs like radio active waste disposal, meteorology, and biological effects of radiation, where considerable work is also beORKBOOK FEATURES
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ing done. Further, commission li censing and inspection procedures are designed to provide protection for the public. A permit to operate a reactor is given by the commission only after a painstaking review of the proposed plant indicates that it is safe. On-site commission inspection of the reactor further enhances safety. Safety Research Reactors
The reactor safety research pro gram includes experimental work with six reactors designed to study the dynamic behavior of reactors when subjected to both sudden and slow changes in reactivity. Three of the reactors are in oper ation (SPERT-I, KEWB, EBR-I) and three are in various stages of construction (SPERT-II, SPERTIII, T R E A T ) . All of these reactors are located at the commission's National Reactor Testing Station, Idaho, except KEWB, at Santa Susana, Calif. SPERT-I (Special Power Ex cursion Reactor Test) is an unpres surized tank type, heterogeneous, water-moderated, thermal reactor. It achieved initial criticality on July 11, 1955. The following description indi cates the experimental procedure employed with SPERT-I to study reactor transients. When the con trol rods are withdrawn, the power increases exponentially. The time in seconds required for the power to change by a factor of e—that is, 2.72—is called the reactor period. When the control rods are rapidly re moved from SPERT-I, reactivity is added and the power can rise on a period as fast as 10 milliseconds or less. In one experiment the power burst reached a peak of 2500 mega watts before this excursion followed the characteristic pattern of selftermination. It is important to em phasize that, in this particular re actor, the power burst was ter minated by the reactor's own inher ent shutdown mechanism and not by the control rods. Objective with SPERT-I is to understand this ki netic behavior and to use this knowl edge to assure safety in future reactor design. Substantial analytical and theoretical programs accompany these experiments. SPERT-Π will be a 300 p.s.i. 400° F. pressurized heterogeneous 86 A
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water reactor designed to obtain ki netic information between SPERT-I, which is unpressurized, and SPERT-III, which is designed to op erate at 2500 p.s.i. and 668 ° F. KEWB (Kinetic Experiment on Water Boilers) is a 50-kw. aqueous homogeneous reactor that has been operating since July 1956. In an aqueous homogeneous reactor the fuel (uranyl sulfate) is in solution, whereas a heterogeneous reactor uti lizes solid fuel plates or shapes. One characteristic of a homogeneous-type reactor is that radiolytic gases are readily produced. These gases pro vide a very effective shutdown mech anism, because voids are created in the moderator which affect thermal neutron production. A recombiner is necessary to eliminate the possi bility of an explosion of the hydrogen and oxygen. KEWB has been used to investigate reactor transients in homogeneous reactors by both slow and rapid removal of the control rods under carefuly planned conditions. From these experiments we are gain ing an understanding of the kinetic behavior of this type reactor. EBR-I (Experimental Breeder Re actor No. 1) is a power reactor exper iment now being used to investigate instability problems associated with fast reactors. A fast reactor is one in which fissions are induced by neu trons which have not been slowed to thermal speeds by a moderator, as is the case in the thermal reactors con sidered previously. A specially de signed core for EBR-I is being studied to determine the stability gained if the fuel elements are held in rigid position and not allowed to bow. The stability effects of series and parallel coolant flow are also be ing investigated. These investiga tions are showing that properly de signed fast reactors can be made very stable. TREAT (Transient Reactor Test) is a new facility being designed by the commission's Argonne Na tional Laboratory. Its principal pur pose is to study the behavior of severe power excursions on various reactor cores and fuel subassemblies to the point of melt down under actual nu clear conditions. Smaller, but nevertheless impor tant, is the in-pile reactor safety pro gram being conducted at the com mission's Oak Ridge National Lab
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oratory. This program is studying the effect of voids introduced into the core or reflector of pool-type reac tors. From a reactor safety point of view it is important to study this phenomenon, since large voids may give rise to large reactivity changes and their sudden collapse could cause a serious reactor power excursion. Metal-Water and Metal-Gas Reactions
Another area of research is the troublesome subject of metal-water reactions. It has been postulated that nuclear excursions could cause the reactor fuel to melt and that the molten fuel and cladding could react exothermically with the water cool ant. Several questions arise in this con nection. First, will the reaction occur in a reactor incident? Second, how much metal will react? And third, what is the rate of reaction and pres sure-time history of such a reaction? We are in the process of finding an swers to these questions. Over the past few years a number of investiga tors have worked on a problem of po tential metal-water reactions follow ing core melt down. Al though a con siderable body of qualitative data has been gathered, the reaction mech anism including rate-controlling fac tors is largely unknown. Both Ar gonne National Laboratory and Gen eral Electric Vallecitos Atomic Lab oratory are working on this problem under contract to the commission. We are somewhat further along with the problem of metal-gas re actions, essentially metal pyrophoricity. Uranium, thorium, and zir conium will apprently ignite at room temperature in bulk form. Investigators at Argonne National Laboratory have approached this problem by separating the variables and by starting the entire investiga tion with good base-line data ob tained by subjecting 1-cm. cubes of pure uranium to isothermal oxidation conditions. In one particular case they observed that a definite break in oxidation rate occurred after about 400 minutes of 50-mm. oxygen at 150°C. In round numbers, the oxy gen consumed changed from 0.1 to 0.6 y per square cm. per minute. This rate change is sharp and well de fined, but unexplained. In other ANL experiments, the
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temperature of the furnace was steadily increased by a programming device. The samples which were in an oxygen atmosphere would, at a certain temperature, suddenly suffer a very rapid acceleration in the oxidation rate. These have been called "burning curves" and from them "ig' nition temperatures" have been reproducibly determined. Addition of 1 atom % of elements like copper and aluminum have a marked effect on ignition time and temperature. Aluminum accelerates and copper inhibits. This excellent work is continuing. Fission Product Release In the important area of fission product release from molten uranium
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reactor fuel, experiments are under way at Oak Ridge National Laboratory. Irradiated fuel samples are melted by induction heating and the volatile fission products are collected and analyzed by radiochemistry. This analysis indicates that the rare gases are the most volatile of the fission products and exhibit the highest per cent of release. As one might suspect, the per cent release depends upon the temperature of the melt and the type of fuel and cladding. In one case with aluminum cladding and uranium fuel, 10% of the krypton and xenon was released in a 700° C. melt. In another case involving zirconium alloy cladding melted at 1850° C , the release was as high as 95%. The halogens, bromine and
Contract Research in Reactor Containment The possible release of nuclear and chemical energy as well as fission products from a reactor make containment of some reactors advisable. Reactor containment is an additional way of protecting the health and safety of the public in the very unlikely event that all preventive measures fail. Location Naval Ordnance Laboratory, White Oak, Md.
Aberdeen Proving Ground, Md.
Stanford Research Institute Armour Research Foundation Argonne National Laboratory Lockheed-Holmes and Narver
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Research Investigating the dynamic properties of the reactor core vessel by both experimental and analytical means. Experiments designed to study the energy absorption of the vessel through the plastic range. Theoretical work consists of developing an equation that would specify the energy absorption of a vessel when subjected to dynamic loading in the plastic range. Analytical and experimental work on the outer, or vapor, containment shell. Explosive charges will be detonated in a series of cylindrical containers, with hemispherical ends. By using vessel diameters of 2l/t, 5, 10, and 20 feet and proper instrumentation, it should be possible to establish scaling laws for full-size containment. It is planned to establish design criteria for containment structures by measuring the elastic and plastic response of these structures and comparing with theoretical predictions. Containment of objects that might be propelled in the event a core vessel ruptures Dynamic loading of structures, theory and design of blast shields, and other containment problems Leak testing of outer containment shell Effects of earthquakes on reactors
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iodine, are also released. Iodine-131 is particularly harmful. Unfortunately, up to 50% of this iodine can be released in the case of zirconium cladding. Reactor Control Another problem is that of reactor control. Although this is a relatively independent field, there is a gray area which impinges upon reactor safety, and the commission is also doing some work in this field. Oak Ridge National Laboratory has a program whose threefold objective is (1) to determine the limit of protection aginst short-period accidents which might reasonably be expected through reactor control, (2) to study safety systems intended by design to attain this limit, and (3) to test components for such systems for various reactor types. A device to keep a reactor from "running away" is highly desirable. To this end, Atomics International under contract to the commission has undertaken research and development on reactor fuses and has tested one model successfully. The idea behind this fuse is to store a reactor poison such as boron trifluoride in the fuse mechanism above the core. When the power level is too high or a power excursion takes place, the poison will be automatically released into a fuse-receiving chamber which absorbs neutrons and shuts down the reactor. Such a fuse is designed to be completely self-contained and will have no connection to outside wires or devices which can be altered. Reactor safety research is not a new undertaking in the commission. It started with the earliest reactors and has grown into a fairly extensive undertaking. Everything is unclassified, in so far as possible. The reactor safety research program is expanding and AEC is making the information available as quickly and extensively as possible, so that reactors can be built both safely and economically.
Our authors like to hear from readers. If you have questions or comments, or both, send them via The Editor, l/EC, 1155 16th Street N.W., Washington 6, D.C. Letters will be forwarded and answered promptly. VOL. 5 1 , NO. 1
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