在核动力装置一回路中使用安全阀

ARTIC LE S USE OF SAFETY VALVES IN THE FIRST LOOP OF A NUCLEAR POWER PLANT WITH A WATER-MODERATED-WATER-COOLED POWER REACTOR F. M. Mitenkov, L. A. Zvereva, B. I. Motorov, E. M. Mel'nikov, and O. B. Samoilov UDC 621.039.587 The rules of domestic (Gosgortekhnadzor, Registr SSSR) and foreign (English Lloyd et ai.) supervisory committees specify that it is necessary to install a safety valve on high-pressure vessels and systems. This requirement is also auto- matically applied to systems in the first loop of nuclear power installations operating under pressure, including those used in ships, nuclear heat power plants, and others. However, the necessity of such a valve in systems in the first loop of a nuclear power plant with a water-cooled-water-moderated reactor is not as obvious as for most high-pressure vessels and systems with a nonnuclear function because of the properties of the nuclear reactor, involving primarily the fact that in order to receive the emitted heat carrier, a special drainage system with biological shielding and its own safety systems (for protecting the overflow tank from rupturing if the safety valve in the first loop is not set) is necessary, since direct emission of radio- active substances into the environment is inadmissable. The opposite case, when the valve fails, i.e., it remains open, differs little from a rupture of the loop. Second, when the heat-calTier discharges through the safety valve, there arises the danger of partial and complete exposure of the active zone and damage to the fuel elements. This sharply increases the activity of the heat carrier and increases the consequences of an accident. Third, when the valve is not set, iv,mediate emergency makeup of the loop is necessary. Thus, the safety valve as a safety device for a nuclear power plant functions only with highly reliable systems for removing heat from the active zone. But, in this case it is possible to prevent the increase in pressure without the valve. In addition, its use increases the length of the pipelines, the number of disengaging connections operating at high pressure, and the quantity of liquid radioactive wastes. The requirements for the reliability of the valve itself are also quite high. We note that in practice cases are known when the valve has failed. Thus, at the Davis Besse nuclear power plant in the USA, the electropneumatic valve in the system of the first loop opened. Instead of lowering the pressure and closing, it completed nine open and close cycles, after which it remained in the open position [ 1 ] . One of the reasons for leakage from the first loop system of the Three-Mile Island nuclear power plant in the USA of a~120 m 3 of heat carrier (more than one-third of the volume) was a failure of the safety valve. In the final analysis, the development of accidents has led to partial exposure of the active zone, damage to the fuel elements, and emission of a large quantity of reactive substances outside the protective jacket. And in other similar installations by the Babcock-Wilcox Company, the safety valve failed repeatedly [2, 3] . Preventing leakage of radioactive substances into the environment with incorrect triggering and especially failure of the safety valve to set is complicated under the conditions of nuclear ship installations by congestion and at nuclear heat power plants by the closeness of population centers. As already noted, when installing a safety valve for safety of nuclear power installations, it is necessary to create special systems which of themselves, or together with the valve, form not so much a system for decreasing express pressure (it depends on the accumulation of excess heat), as systems for removing heat from the active zone with sufficient operating time independent of external energy sources and controlling actions. The same problems also occur in the development of the basic heat removal systems, and therefore, they can be solved by the same methods. (An increase in pressure can also be caused by external reasons, e.g., by a fire. But even in this case, the problem reduces to creating a system for removing heat both in the presence of a safety valve, as well as without it). The problem of preventing an increase in pressure was solved in this manner when designing domestic shipborne nuclear power plants. In what follows, we examine reliable methods for protecting the first loop of the nuclear power plant from pressure increases above an allowable level. It is useful to begin the discussion with possible (and hypothetical) accidents. Accidents That Can Lead to an Increase in Pressure. An increase in pressure in the loop as a result of an accumula- tion of excess heat is possible in accident situations due to cessation of circulation of heat carrier in the first loop and injection of feedwater into the steam generators, increase in the reactor power above the allowable level, a fire, as well as improper operation of the makeup system for the loop. In its turn, the reasons for cessation of feedwater delivery could include disconnection of the nuclear power plant, malfunctioning of the pump, a break in the pipeline for the feedwater, spontaneous closure of the regulating valve, etc. Increase in the reactor power is caused by displacement of the control rods, overshooting of the cold heat carrier with startup of pump for a previously cutoff loop, spontaneous increase in the t Translated from Atomnaya Energiya, Vol. 50, No. 5, pp. 308-3t0, May, I981. 0038-531X/81/5005- 0275807.50 �9 1981 plenmn Publishing Corporat ion 275 flow rate of the feedwater or flow rate in the first loop (self-adjustment effects). It is evident that these primary accidents can also originate from external reasons both of a general nature (flooding, running aground, or capsizing), as well as the failure of a specific element in the structure or automatic systems, operator error, etc. The safety code for nuclear-powered merchant vessels, developed by the International Maritime Consulting Organiza- tion, does not contain a clearly fixed number of and list of analyzable independent failures. According to the domestic "Rules for Safety of Nuclear Electric Power Plants" [4], the analysis was carried out for the following conditions: the initial accident (first failure) involves a defect (previously not discovered over a long period of time) in the equipment indirectly carrying out shielding or localizing functions; one of the independent active shielding systems and one of the independent active localizing systems also failed or malfunctioned. Methods for Preventing Accidents. A malfunction of the makeup system for the first loop is easily avoided by installing a safety valve built into the makeup pump, which guarantees that the pressure in the makeup pump will not exceed the allowable level. Naturally, in this system, the safety vane is free of the dangers characteristic of the radioactive medium in the first loop. Accumulation of excess heat in the first loop is prevented as follows: redundancy in the system for protecting against accidents, control, and signaling according to the degree of importance of the shielding barrier (confinement, suppression, warning, slow and operational shielding); redundancy as an assurance of shielding actions according to deviation of several dependent and cause-and-effect coupled parameters (turnover of circulation pumps, level of circulation, flow rate of feedwater, power, temperature, pressure, loss of electrical power, malfunctions in automatic systems, etc.); multichannel nature of measuring circuits; independent panels, paths, and sources for electrical power distributed over some area; auxiliary pipelines for injecting feedwater into the cooling system; the presence of several shutdown cooling channels, tanks for storing water, and other means. Thus, the basic principle for constructing a highly efficient system for accident protection, in particular for protecting a reactor from accidents, is the creation of several protective barriers by doubling them sequentially, e.g., when feedwater injection ceases, the temperature and pressure increase in addition to the self-regulation effects in the reactor. Preventive action automatically decreases the power to the point at which the deviation of a parameter above a given value is compensated. Special measures provide for the case when the nuclear power plant is cut off: independence of the accident protection system from electrical power sources (placement in operation by spring action); providing for a level of natural circulation in the first loop sufficient for removing the residual heat liberated; feeding water into the cooling system in the first Ioop by a hydropneumatic accumulator. Inactivation of the regulating elements for any position of the vessel in space is avoided by the structural design of the drive. Examples of the Most Important Accidents Taking into Account Superposition of Failures. 1. Cessation of electrical power flow from the turbogenerators of the nuclear power installation is accompanied by malfunction of one of the standby diesel generators and switching-off of the second one due to overloading. Provisions are made for switching-on reliably a reserve, less powerful, diesel generator with a reduced number of users. 2. Cessation of feedwater injection from the main feed pump with the shutdown cooling channel disconnected simultaneously through the cooler filter or malfunction of the system providing for low speed in three of the circulation pumps. Provision is made for injecting feedwater from the standby feed pump along a separate pipeline, bypassing the condensor feeding system. 3. Spontaneous opening of the feed valve with the warning signals and emergency protection signals for excessive power not getting through at the same time. Safe passage through this regime is provided by the temperature and pressure warning and emergency shielding systems in the first loop. 4. Overshooting of cold water from the steam generator into the active zone when a disconnected loop is turned on due to leaky cutoff valves. The concomitant malfunctions include nontriggering of power warning and emergency shielding systems and malfunctioning of the feedwater heater. This regime is provided for by warning and temperature and pressure warning safety systems. An additional organizational measure is disallowing direct startup of pumps to a high speed. In all regimes of this type, the pressure in the first loop of the nuclear power plant in the icebreaker Lenin did not exceed 276 P, MPa 19 18 f? lg I5 14 _ I I I I _ I lO00 2000 YOOO 4090 50gOt, see Fig. 1. Pressure in the first loop of the nuclear power plant as a function of time with compartments flooded at 1000 sec. the allowable value. Figure 1 shows the change in pressure under the condition that the flow rate of the feedwater decreases !inearly over 10 sec from the nominal value to zero; heat is liberated into the environment through air gaps over a period of 1000 sec; after 1000 sec, everything is filled with water and heat transfer directly into the cold water begins. As prolonged successful experience in the use of icebreakers has shown, discontinuing the use of a safety valve in the first loop does not decrease but increases the safety of the nuclear power plant. Conclusions. It is not useful to transfer the requirements of the supervising committees on the safety of the usual vessels and high-pressure systems against their exceeding the allowable value of the pressure in nuclear installations with the conditions of radioactivity in the first loop of nuclear power plants, especially those onboard a vessel, with such a system as the active zone, which does not permit exposure and prolonged liberation of heat. Protection against an increase in pressure can be more effectively provided by reliable systems for shutting down the reactor, removing heat, and others, which is indicated by the comparative experience in using nuclear-powered icebreakers and some nuclear power plants abroad. LITERATURE CITED 1. B. Verna, Nucl. News, 22, 34 (1979). 2. Personal Reflections on the Kemeny Report, Nucl. Eng., 21, 75 (1980). 3. "Not one, but two LOCAs at Three-Mile Island," Nucl. Eng. Int., 24, 10 (1979). 4. Rules for Nuclear Safety of Nuclear Power Plants [in Russian], PBYa-04-74, Atomizdat (1977). 277