Heat transport system

Rapid bolt-up and fill-up provisions for the heat transport system to allow the steam generators to function as a heat sink in the event of problems with the shutdown cooling heat sink when the heat transport system is drained to the header level... 

The essential ingredients for producing heat in a thermal fission nuclear reactor are the fuel and a moderator. A heat transport system with its coolant is necessary to convey the heat from the reactor to boilers where steam is produced to drive the turbogenerator. The natural materials available for fuel and moderator are uranium ore and water natural uranium extracted from the ore comprises the fissionable isotope uranium-235 and water contains hydrogen which is a good moderator. (Table I)... 

During normal operation, the main circulator transports hot helium at 1266°F (686°C) from the bottom of the core to the steam generator which, in turn, produces superheated steam at I005°F (541 °C) and 2500 psia. The cold helium at 496°F (258°C) is returned to the top of the reactor core. During normal shutdown and refueling, the non-safety auxiliary shutdown heat removal system removes core afterheat if the main heat transport system is not operational. 

Table 2 Summary of hydrogen PIRT chart for heat transport systems...

A common thread in many of the reactor technologies that currently exist or that are under development is the use of water as the heat transport medium (the coolant ). In many respects, water is an ideal coolant, because it has a high heat capacity, can be obtained in a high purity, is inexpensive, has a wide liquid range (0-374.15 °C), is easily handled, and had been used since the dawn of steam power. Thus, in their most fundamental form, water-cooled nuclear reactors (WC-NRs) comprise a nuclear boiler, a heat transport system (piping, channels, steam generators, etc.), a set of turbines (high pressure, intermediate pressure, and low... 

Suzuki, T. et al., 1996, Monju Secondary Heat Transport System Sodium Leak , Proceeding of 1 0 Pacific Basin Nuclear Conference, Kobe, Japan. 

The capital cost of FBR is about 1.5 to 2.5 times that of thermal reactors and significant cost reduction is essential for its successful deployment. Cost reduction measures adopted in LMFBR include elimination of ex-vessel sodium storage, decrease in number size of components of heat transport system, compact layouts, increasing operating temperature, increasing plant life and increasing fuel bumup. 

As mentioned above, elimination of all feedback control systems from the reactor and secondary heat transport systems makes the 4S plant control system very simple and economic. 

When the reactor is shut down for maintenance or refueling, decay heat can be removed from the core by the normal Heat Transport System (HTS) described above, or alternatively by an independent Shutdown Cooling System (SCS). The SCS consists of a motor-driven circulator coupled with a water-cooled heat exchanger mounted beneath the reactor core within the reactor vessel. The SCS is provided for investment protection and flexibility of operation. The SCS and HTS are not "safety-related". 

The primary components of each RS (core, reflectors, and associated supports, restraints, and controls) are contained in the reactor vessel. The nuclear heat is generated in the reactor core. Removal of the heat energy is provided by the Heat Transport System (HTS) with the main circulator providing the driving force to supply helium coolant into an upper core inlet plenum and to draw heated coolant from a bottom core outlet plenum. The primary coolant is distributed to numerous coolant channels running vertically through the core. The outlet plenum directs the flow to the central portion of the coaxial cross duct which channels the helium flow to the steam generator vessel (see Chapter 5). 

DBE-2 is a Heat Transport System (HTS) transient without control rod trip, requiring reactor shutdown with the reserve shutdown control equipment (RSCE), and core cooling with the Shutdown Cooling System (SCS). During the first 5 minutes of the event the core temperatures rise by a maximum of about 80 C (150°F) above the normal operating levels. After this initial phase, the core gradually cools down. 

On Etecember 8, 1995, a leakage of sodium occurred in the piping room (C) of the Secondary Heat Transport System (SHTS) while the output of the reactor was being raised for a plant trip test at 40% output as part of a series of performance tests. The nuclear reactor was shut down manually after the accident, and sodium was drained from the SHTS in which the accident occurred and also from the Loop C of the Primary Heat Transfer System (PHTS). The plant is currently in a low-temperature shutdown state. The plant conditions of Monju at the time of the sodium leak occurrence are shown in Fig. 3.1. 

PHTS Primary Heat Transport System SHTS Secondary Hert Transport System... 

Safety evaluation studies have been conducted for confuming the physical phenomena and integrity of the fuel subassemblies, the core internal structures and the heat transport systems during the normal operation, scram transients and the early stage of postulated accidents. On this account, thermohydraulic experiments related to the decay heat removal by natural circulation have been carried out, and the development and validation of the thermohydraulic safety analysis codes is also in progress. 

Another ongoing item on the steam generator safety is a leak detection system development for the double wall tube that is being developed for the steam generator installed in a primary heat transport system of a future FBR plant. Basic performance data was obtained for an outer tube leak detection system to detect helium leak into sodium. An inner tube leak detection system using infrared rays is also under developing to quickly detect humidity at gas plenums. 

In fact, the sodium leak that occurred at Monju involved the secondary heat transport system, which means that the leaked sodium was not radioactive. Therefore, no radioactive material was released to the environment by the sodium leak and the safety facilities were not affected. Knowledge about sodium fire has been accumulated through operation experiences. So far, there have been about 150 sodium leak events in the UK, France, Germany, and the former Soviet Union. However, no reports have been released of safety facility damage due to the sodium leak or combustion, and the reactors that experienced the sodium leakage were repaired and successfully resumed their operations. 

Small reactors oHitain less stcxed energy and often cerate at a lower pressure in the primary heat transport system than typical large pow reactors. In addition, they have lower core power density and passive systems or features instead of active (Hies for heat tran xat and control. These characteristics are basically conducive to operation without fully lic ised operat(HS on-site. Some of the prerequisites for such operation are ... 

From thermodynamic point of view, the higher enthalpy will give higher efiBciency and to do so, normally the thermodynamic conditions of the working fluid is translated directly to the op>erating condition of the primary loop of heat transport system. 

Early studies concluded that the nuclear heat transport systems of a small reactor could not operate significantly above atmospheric pressure, if all design requirements were to be satisfied. It was noted that all of the very small reactors currently available, and generally regarded as inherently safe (SLOWPOKE, TRIGA, MAPLE-X) are pool type reactors. 

The Double Pool concept has been driven by an objective to reduce fast reactor construction costs to the same level as those of an LWR. A major contribution has been to reduce the overall size of the intermediate heat transport system by installing the steam generators in the sodium filled annular space between the primary vessel and the guard vessel. These two examples serve to show some of the ways in which the basic modular concept can be modified to meet different objectives. 

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