12.187. The destruction of Unit 4 of the Chernobyl Atomic Power Sta­tion in Ukraine on April 26, 1986 was the worst nightmare of nuclear engineers throughout the world. Attempts to convince a skeptical public that nuclear power is indeed a safe energy option would now be more difficult than ever, even though the reactor design is unique to the former Soviet Union and inherently unstable. A first step in discussing the accident and its impact is to become familiar with the system.

12.188. The reactor, one of the RBMK-1000 type, was boiling light — water-cooled and graphite-moderated. Fuel element subassemblies, within pressure tubes through which the coolant flowed, contained 18 rods ar­ranged in concentric rings. At full power, the coolant channel exit steam quality was 14 percent. The core, 7 m high and 11.8 m in diameter, con­tained 1661 such channels. Refueling was accomplished at full power by a machine in a bay above the core, as shown in the Fig. 12.14 plant arrange­ment. A 2 percent enrichment was used, with 211 control rods of various types needed for reactivity compensation. The rated thermal power was 3200 MW.

12.189. As a result of the massive size of the reactor and fuel handling equipment, it was considered impractical to have a complete containment such as used for all U. S. reactors. Instead, an accident localization system, as indicated in Fig. 12.14, was provided. This consisted of various sealed compartments that enclose the circulating pumps, large piping, and other components, the failure of which could lead to a LOCA. This compartment system vents to a suppression pool. The cooling channel refueling con­nections were unprotected. Hence, the accident localization system was designed for a different type of accident, i. e., a large pipe LOCA, rather than the massive fuel channel failure that occurred.

12.190. Since graphite is the moderator in the RBMK-1000 reactor, voiding of the coolant, which is a neutron absorber, increases the reactivity. This positive reactivity coefficient was significantly greater at low power. Therefore, 700 MW(t) was specified as the minimum permissible contin­uous operating power level. An interesting positive reactivity effect is in­troduced when control rods are inserted as a result of the displacement of water at the bottom of the control rod channels by graphite control rod

Primary coolant boundary areas not protected by AtS Semi protected mm

Protected areas

followers. During post-accident analysis, it was also concluded that the emergency rod insertion rate was slow by western standards.

12.191. The accident occurred during a test of turbine-generator coast­ing-down power, which was used to drive an emergency feedwater pump for about 1 minute in the event of the loss of off-site power, with some of the desired electrical conditions simulated. Although the intended reactor power for the test was 700 MW(th), errors by the operators resulted in a power loss to 30 MW(th), where xenon growth, particularly at the bottom of the core, from the previous higher power operation made it difficult to increase the power without withdrawing almost all of the control rods. Even with such action, only a power level of 200 MW(th) could be achieved, a level in violation of operating procedures because of the inherent insta­bility of the reactor.

12.192. The test plan was initiated at this low power level by starting additional recirculating pumps as called for by an electrical simulation, which had the effect of reducing core voids and causing additional control rod withdrawal. As a result of operating difficulties, various protection devices were blocked out by the operators. When the recirculating pumps were allowed to coast down, as planned, coolant flow decreased and voids re-formed very rapidly in the pressure tubes, which increased reactivity because of the positive void coefficient, particularly at the bottom of the core. An emergency scram (trip) was initiated manually almost immedi­ately, but the almost fully withdrawn rods could not be inserted fast enough to prevent a prompt critical power excursion. In fact, the rod insertion introduced some additional reactivity at the core bottom as a result of water displacement by the graphite followers.

12.193. The rapid vaporization of the coolant in the pressure tubes generated a shock wave that ruptured most of the tubes. Apparently, there were two excursions, seconds apart, the second a result of almost complete coolant voiding. The fuel became molten and generated an immense quan­tity of steam, which blew the 1000-ton steel and concrete biological shield off the top of the reactor. Hydrogen, formed by the reaction of fragmented cladding and water, exploded, severely damaging the building. Pressure tube ruptures provided an inlet and an outlet for air to feed combustion of the graphite, which was probably ignited by the exothermic zirconium — niobium oxidation reaction. The fire continued for several days and cer­tainly complicated the management of the accident. The accident was cat­egorized as level 7 on the International Nuclear Event Scale (INES).[25]

12.194. In late 1986, after radiation levels had decreased somewhat, the damaged reactor building was enclosed by a concrete and steel shell, or “sarcophagus.” Several years later, explorations inside the sarcophagus revealed that about 96 percent of the original fuel is contained in solidified “lava flows” in chambers below the reactor vault and in the form of dust and particles distributed inside the building. Since it was necessary to construct the sarcophagus as an emergency measure upon damaged foun­dations, it is expected that replacement will eventually be necessary.

12.195. The total release of particulate fuel from the core is estimated at 3.5 percent of the original inventory. This corresponds to a radioactivity release into the environment on the order of 50 million curies. Fallout over parts of the former Soviet Union and other countries was widespread. Considering the most biologically sensitive fission products, 100 percent of the rare gases was released, about 20 percent of the iodine, and roughly 13 percent of the cesium. The estimated 2 million curie release of the 30- year half-life cesium-137 is the most significant long-term contamination contribution. The initial 32 fatalities all occurred on-site. The long-term consequences of the exposure received by about 200 plant workers treated for radiation sickness at the time of the accident and off-site exposure to fallout by some of the nearby general population remains uncertain.