Water-cooled reactor, accidents

Vessel blowdown. The previously mentioned relationships for the critical flow rate of a steam-water mixture can be employed with the conservation of mass and energy for a vessel of fixed volume to determine its time-dependent blowdown properties. The range of problems associated with coolant decompression in water-cooled reactors is quite broad. The types of hypothetical (some are even incredible) reactor accidents may be... 

At the present state of the art, a corner of an article about evaluation of population hazards is hardly an appropriate place in which to attempt an exposition of reactor safety. Nevertheless, we may contrive a brief description of these types of reactor accidents which, it is thought, could lead to fission product release. The intention is to illustrate ways in which fuel could be damaged and then release fission products ultimately to the atmosphere. Though gas-cooled reactors, water-cooled reactors, and sodium-cooled fast reactors will be discussed, no comparisons, invidious or otherwise, are intended between the safety of these systems. 

The Chernobyl accident, as well as that at Three Mile Island, therefore suggests that the circumstances under which a rapid rise in cladding temperature can occur should receive special study, both experimentally and analytically, for water-cooled reactor systems. 

Water cooled reactors. 2. Boiling water reactors. 3. Nuclear reactor accidents. 4. Nuclear power plants — Accidents. 5. Nuclear reactors — Safety measures. 6. Light water graphite reactors. I. International Atomic Energy Agency. II. Series. 

Postma, A.K. and Zavadoski, R.W., Review of Organic Iodide Formation Under Reactor Accident Conditions in Water-Cooled Reactors, Atomic Energy Commission Report, WASH-1233, (1972). 

ANSI/ANS 4.5, "Criteria for Accident Monitoring Functions in Light-Water-Cooled Reactors", December 1980. 

Amey, M, D. H, Johnson, P. A. V. Improved chemistry Key to reduce occupational radiation exposure in water-cooled reactors. Nucler Europe Worldscan X, 34-36 (1990) Anthoni, S., Ridoux, P, Chevalier, C. Evaluation of corrosion products in a pressurized water reactor during shutdown. Proc. 5. BNES Conf. Water Chemistry of Nuclear Reactor Systems. Bournemouth, UK, 1989, Vol. 2, p. 35-39 Antipov, S. A., Dranenko, V. V., Konkov, V. F., Sokolov, I. B., Khaikovskij, A. A. Influence of fuel clad surface on performance under operational and accident conditions. Report... 

Postma, A. K., Zavadoski, R. W. Review of organic iodide formation under accident conditions in water-cooled reactors. Report WASH-1233 (1972)... 

Essential equipment here means eqnipment that mnst continue to be operable to prevent the escalation of an accident or further radioactive releases (e.g. pumps in water cooled reactors or gas circulators in OCRs, which are required to maintain core cooling), and equipment that is required for monitoring the state of the plant after an accident. 

As a coolant operating at atmospheric pressure, the loss of coolant accident is virtually eliminated by use of an appropriately designed guard vessel. This is not only a safety advantage, but also offers additional potential for plant simplification and improved economic performance, since the complex process of simultaneous management of temperature, pressure, and coolant level (as is seen in water-cooled reactors) is not necessary. 

Dang, G., Huang, D., Lu, J., et al., 2013. Large-break accident analysis of supercritical water-cooled reactor CSRIOOO using APROS code. Nuclear Power Engineering 1, 72-82. 

The potential applications of SiC SiC composites currently considered are core components, especially the control rod sheath and cladding of the VHTR, GFR, SFR, MSR, and LFR. Because the scope of this chapter is restricted to the Generation IV system, only part of the nuclear applications of SiCf/SiC was described. However, SiC SiC composites are also considered as the in-vessel components of magnetic confinement fusion devices including blanket structures, flow channel inserts (FCI) for the liquid metal (LM) blankets, and plasma-facing components (PFCs) [88—91]. In addition, they are candidates for an advanced fuel cladding for LWRs as an ATF (accident tolerance fuel) concept [72,92—97] and a channel box for the BWRs (boiling water-cooled reactors) [96,98,99]. 

From the beginning of the conceptual study on supercritical water cooled reactors, several plant transient analysis codes have been developed, modified, and applied to them [1-9]. The general name of these codes is Supercritical Pressure Reactor Accident and Transient analysis code (SPRAT). SPRAT mainly calculates mass and energy conservations, fuel rod heat conduction, and point kinetics. The relation among these calculations is shown in Fig. 4.1. SPRAT can deal with flow, pressure, and reactivity induced transients and accidents at supercritical pressure. The flow chart is shown in Fig. 4.2. 

Nuclear power plant systems may be classified as "Frontline" and "Support. . iccurding to their. service in an accident. Frontline systems are the engineered safety systems that deal directly with an accident. Support systems support the frontline systems. Accident initiators are broadly grouped as loss of cooling accidents (LOCAs) or transients. In a LOCA, water cooling the reactor is lost by failure of the cooling envelope. These are typically classified as small-small (SSLOCA), smalt (SLOCA), medium (MLOCA) and large (LLOCA). 

The first three submarine reactors disposed of in this way seems have originated from a Hotel, a November and a Yankee class submarine, all of early designs, and they had all suffered either a criticality or a loss-of-cooling accident. They were all provided with two pressurized water reactors. The submarine was the Project 645 submarine with a November class hull, which was provided with two liquid-metal cooled reactors. It had suffered a loss-of-cooling accident in one of its reactors [4]. 

Safety features at a nuclear power plant include automatic shutdown of the fission process by insertion of control rods, emergency water cooling for the cote in case of pipeline breakage, and a concrete containment shell. It is impossible for a reactor to have a nuclear explosion because the fuel enrichment in a reactor is intentionally limited to about 3% uranium-235, while almost 100% pure uranium-235 is required for a bomb. The worst accident at a PWR would be a steam explosion, which could contaminate the inside of the containment shell. 

Decay heat in fuel elements is assumed to be dissipated by means of heat conduction and radiation to the outside of the reactor pressure vessel, and then taken away to the ultimate heat sink by water cooling panels on the surface of the primary concrete cell. Therefore, no coolant flow through the reactor core would be necessary for the decay heat removal in loss of coolant flow or loss of pressure accidents. The maximum temperature of fuel in accidents shall be limited to 1 bOO C. 

As explained in Section II, in a depressurization accident to a gas-cooled reactor (and to some water reactors, e.g., the steam generating heavy water reactor—SGHWR) there is a possibility that can temperatures might rise to about 1000 C at which point some cans could fail through their internal gas pressure. Laboratory experiments in which pre-irradiated fuel elements... 

Accident evaluations specific to the GT-MHR confirmed that the passive safety characteristics of the previous steam cycle modular high temperature gas-cooled reactor designs were maintained. Events initiated by one or more turbine blade failures were assessed. It was found that the resulting differential pressure forces across the prismatic core did not exceed the allowable graphite stresses. Since the dominant risk contributor for the steam cycle design were initiated by water ingress from the steam generators, the GT-MHR is expected to have a lower risk profile to the public. References 4 and 5 provide more information on the GT-MHR safety evaluations. 

Because the HTR-10 test reactor is designed on the inherent safety philosophy, safety classifications of systems and components departure from the way it is done for water cooled power reactors For example, primary pressure boundary is defined to the first isolation valve Steam generator tubes are classified as Class II component Diesel generators are not required to be as highly qualified as those used for large water cooled power reactors, since no systems or components with large power demand would require an immediate start of the diesel engines at a plant black-out accident... 

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