Superheated Helium

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. 

As was discussed in Section 3.3.1, the theoretical value of the index, y, is 1.67 for a monatomic gas such as helium or argon. 1.4 for diatomic gases such as hydrogen, oxygen and nitrogen and 1.33 for a polyatomic gas such as carbon dioxide or superheated steam. The true indices will deviate somewhat from these values in practice for instance the value of 1.3 is normally used as a better approximation for superheated steam. 

In the earlier phases of breeder reactor development, especially in the 1950s and 1960s, high pressure gases, such as helium,C02 or superheated steam were studied. Between 1960 and 1970, H2 0-steam cooled and D2 0-steam cooled fast reactor concepts were studied in the USA and the former FRG. Helium cooled fast reactor concepts have been pursued as an alternative coolant concept in Europe and the USA. Some fuel development for a CO2 cooled fast breeder has been continued on a small scale in the UK. Lead-bismuth alloy as a coolant was studied in the former USSR for propulsion and land based reactors. 

The main system arrangement is shown in Fig. 2. The high temperature secondary helium gas flows at first into the steam reformer and then into a superheater and a steam generator. The superheated steam is provided mainly to the steam reformer. 

The primary coolant circuit was wholly contained within the PCRV with the core and reflectors located in the upper part of the cavity and the steam generators and circulators located in the lower part. The helium coolant flowed downward through the reactor core and was then directed into the reheater, superheater, evaporator, and economizer sections of the 12 steam generators. From the steam generators, the helium entered the four circulators and was pumped up around the outside of the core support floor and the core barrel before entering the plenum above the core. The superheated and reheated steam was converted to electricity in a conventional steam cycle power conversion turbine-generator system. 

Substances considered in a compilation of the thermodynamic properties of refrigerants include hydrogen, parahydrogen, helium, neon, nitrogen, air, oxygen, argon, carbon dioxide, hydrocarbons (e.g. methane, ethane, propane, butane, isobutane, ethylene, and propene), and fluoro-and fluoro-chloro-hydrocarbons. Properties listed include those for the liquid and saturated vapour, superheated vapour, and unsaturated vapour. In addition, pressure-enthalpy, and in some instances pressure-entropy, diagrams are provided. 

The arrangement of the primary circuit of the reactor is illustrated in Fig. 8.5. The cooling system is divided into two loops, each of which has a six-module steam generator and two steam-driven helium circulators. The gas from all four circulators discharges into a plenum underneath the core support floor. The full flow then passes up the outside of the core to the core inlet plenum above the core. It then flows downward through the core, where it is heated to a temperature of 780°C, to the steam generators, to produce superheated and reheated steam. The steam turbines of the circulators are... 

The HTR-PM shown in Fig. 3.9 contains two parallel trains of nuclear steam supply system (NSSS) of identical design, each consisting of a 250-MWth pebble bed reactor and a steam generator. The two NSSS systems have independent primary loops but share auxiliary facilities, such as fuel handling system and helium purification system. The two trains jointly supply superheated steam to a common steam turbine power generator rated at 200 MWg. 

J. A. Luker and Thomas Gniewek, Saturation Composition of Steam— Helium—Water Mixtures PVT Data and Heat Capacity of Superheated Steam— Helium Mixtures, USAEC Report AECU-3299, Syracuse University Research Institute, July 29, 1955. 

I hi systems reciuired in the plant are shown by the flowsheet of Fig. 25-4. I he hi at is removed from the reactor by helium at 500 psia, which leaves the reactor at 1300° and returns at 900°F. 4 his heat is removed from the helium in a steam generator that produces superheated steam at 8.50 psig, 900°F. The. steam is utilized by a standard turbine generator plant. 

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