Nuclear heat supply system

The nuclear heat supply system generates heat in the reactor core and produces steam in the IHTS. The IHTS SG is composed of a helical coil counterflow heat exchanger, startup recirculation tank and pump, leak detection system, and isolation valves. Intermediate sodium flows on the shell side of the helical coil tubes, through which water flows to produce superheated steam. Two reactor SGs feed one turbine island to form a power block. 

The Nuclear Steam Supply System based on the standard Westlnghouse four-loop pressurised water reactor produces 3425 MW of heat. 

The CANDU 3 nuclear steam supply system includes the reactor, the heat transport system, the moderator system, the shutdown cooling system, the fuel handling system and associated control and service systems located within the reactor building (Figure 5.8.1 A) and the emergency core cooling system. The followng sections provide detail on specific NSSS system. 

The heat produced by controlled fission in the fuel is transferred to the pressurized heavy water coolant and circulated through the fuel channels and steam generators in a closed circuit. In the steam generators, the heat is used to produce light water steam. This steam is used to drive the turbine generator to produce electricity. The nuclear steam supply system (NSSS) is illustrated in Figure 4.2. 

PRISM employs passive safety, digital instrumentation and control, and modular fabrication techniques to expedite plant construction [1-4]. PRISM has a rated thermal power of 840 MW and an electrical output of 311 MWe. Each PRISM module has an intermediate sodium loop that exchanges heat between the primary sodium coolant from the core with water/steam in a sodium-water steam generator (SG). The steam from the sodium-water SG feeds a conventional steam turbine. A diagram of the PRISM nuclear steam supply system (NSSS) is shown in Figure 6.2. 

This corresponds to an exit temperature of 622°F. The nuclear sfeam supply system (NSSS) thermal output is the heat transferred in the steam generator, which in this case equals the reactor thermal power and the energy added in the pump (roughly 3415 MW). The required steam flow rate on the secondary side is calculated by Equations 23.14 and 23.69 where is neglected. 

In the present chapter, we shall first consider the details of the heat generation process within the reactor core. The transfer of heat from the fuel elements to the coolant will then be covered for both the nonboiling and boiling cases. Finally, the overall design of the nuclear steam supply system (NSSS) will be discussed. 

Figure 1-2 shows the simplified schematic diagram of the SMART nuclear steam supply system (NSSS) and exhibits the safety systems and the primary system as well as auxiliary systems. The engineered safety systems designed to function passively on demand consist of a reactor shutdown system, passive residual heat removal system, emergency core cooling system, safeguard vessel and reactor overpressure protection system. 

The heat supply system from a nuclear cogeneration plant includes ... 

With this target in mind, a plant design concept was established for JSFR. It is a loop-type plant with a two-loop heat transport system. Designs for a commercial version with 1500 MWe and a demonstration version with 750 MWg are pursued in the design study. A bird s eye view of the nuclear steam supply system (NSSS) of JSFR design is illustrated in Fig. 11.1, and the major design specifications are summarized in Table 11.1 for the demonstration version design. 

KALIMER is a pool-type liquid metal-cooled reactor that has four intermediate heat exchangers (IHXs), four electromagnetic-type primary coolant pumps. In the KALIMER conceptual design [3], focus has been on the nuclear steam supply system (NSSS) and essential BOP (Balance of Plant) systems. The ultimate objectives for the KALIMER conceptual design are to make it safer, more economical, more resistant to nuclear proliferation, and yield less impact on the environment. Figure 1 represents the sehematie of the KALIMER NSSS. 

Fundamental to all heat-power conversion systems is that a significant portion of the heat supplied to the system must be rejected. Depending on their heat-power conversion efficiency, fossil-fueled plants waste 40-60%, nuclear-fueled plants 60-70% of the heat input geothermal power plants make no exception here. 

In the case of solar powered systems, decoupling of heat source and chemical plant facilitates the compensation of fluctuating and intermittent available power input. This is particularly important if units of the chemical system require steady-state conditions over long periods. Beyond that decoupled systems allow for an easier integration of thermal and chemical storage units to compensate daily or seasonal variation of solar supply. The same applies for hybrid operation, i.e. the combination with burner firing or with a nuclear heat source. 

The new system jointly developed by SSC RF-IPPE (with its subsidiary ECS-Russia), DAO PROMGAZ of RAO GAZPROM and RUHRGAS AG is based on a TEG design, which proved successful as a component of nuclear power plant and has to date the experience of stable operation for not less than 18 years with not less than 1300 thermal cycles (on/ofif). By now it is has been modernized by the development participants for use in natural gas heated systems and equipped with adchtional devices for the gas supply, the control over performance of power units and power/heat recovery system, the functioning remote control etc. 

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). 

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