Pressurized water reactors secondary loop

Most nuclear reactors use a heat exchanger to transfer heat from a primary coolant loop through the reactor core to a secondary loop that suppHes steam (qv) to a turbine (see HeaT-EXCHANGETECHNOLOGy). The pressurized water reactor is the most common example. The boiling water reactor, however, generates steam in the core. 

Men et al. (2014) have conducted experiments on the natural convection heat transfer for a PRHRS HEX in an in-containment refueling water storage tank. Several empirical correlations for the forced convection flow internal to the HEX tube and the natural convection heat transfer outside of the tube in the tank, for the vertical and horizontal portion of the tube, were compared with experimental data. The Dittus-Boelter forced convection correlation and the McAdams correlations for natural convection proved to give the better model of the data. Wenbin et al. (2014) have conducted experiments for the secondary loop of the Chinese Advance Pressurized Water Reactor for validation of the MIS AP20 models and code. These and other papers are in a special issue of the Science and Technology of Nuclear Installations journal published in 2014 as indicated by the cited references. 

In the other design, PWRs have two closed loops of water circulating in the plant plus a third, external loop to remove the waste heat. Water is pumped through the reactor core in the primary coolant loop to moderate the neutrons and to remove the heat from the core as in the BWR. However, the reactor vessel is pressurized so that the water does not boil. Steam is necessary to run the turbines, so the primary loop transfers the heat to a secondary loop. The water in the secondary loop is allowed to boil, producing steam that is isolated from both the core and the outside. The water in the primary loop usually contains boron (as boric acid H3BO3 0.025 M) to control the reactivity of the reactor. The steam in the secondary loop is allowed to expand and cool through a set of turbines as in the BWR the cold steam condenses and is returned to the primary heat exchanger. A third loop of water is used to maintain the low-temperature end of the expansion near room temperature and remove the waste heat. 

Another method for dealing with high reactor temperatures is to generate steam, as shown in Fig. 4.19. Here we allow the coolant to boil and thereby provide a constant jacket temperature. The secondary loop controls pressure in the boiler drum by venting steam. Fresh boiler feed water is added by level control. A potential problem W ith this arrangement is the possibility for boiler swell that results in an increase in the level due to increased vaporization in the jacket. The increased level due to swell reduces the intake of boiler feed water when in reality it should be increased. This problem can be overcome by providing a ratio controller between the steam flow and the feed water with the ratio reset by the steam drum level controller. Boiler feed water flow will now change in the correct direction in response to load. 

Energy produced in the reactor is carried away by means of a coolant such as pressurized water, liquid sodium, or carbon dioxide gas. The circulating coolant absorbs heat in the reactor once outside the reactor, it is allowed to boil or the heat it contains is used to boil water in a secondary loop. Steam produced in either of these ways is then piped into the electrical generating unit, where it turns the blades of a turbine. The turbine, in turn, turns a generator that produces electrical energy. 

The sodium in the secondary loop takes heat from the primary sodium via an intermediate heat exchanger, in it is heated to a temperature of 505X from an inlet temperature of 325°C. Next, heat from the secondary loop is transferred to water in a helical coil-type steam generator system which consists of an evaporator and a superheater. Superheated steam with a pressure of 12.7 MPa and a temperature of 483 C is injected to a turbine directly connected to a generator. The thermal output of the reactor is 714 MWt and the electrical output 280 MWe. 

There is a capability to isolate each primary loop from the reactor using two gate valves on the suction and pressure pipelines of the circuit. On the pressure pipeline of each loop downstream of the PSP a flap-type check valve is provided eliminating coolant backflow in the event of a PSP trip in one loop when the other PSPs are operative. The secondary sodium circuits comprise EHX heat transfer tubes, pipelines, secondary sodium pumps and steam generators. Due to utilization of the reactor energy for fresh water production the steam-water system has some specific features. Steam from the SG is supplied to turbines of two types a condensing turbine (K-100-45) and a back-pressure turbine (K-50-45). Exhaust steam flows from the back-pressure turbine and from intermediate bleeds of K-100-45 turbine are supplied to the water desalination facilities. At a heat output of 750 MW the reactor produces ... 

During reactor operation, the secondary system is operated as a steam condensate system to transfer heat removed from the primary loop to the circulating raw water system. Heat transfer from the primary loop results in boiling on the secondary side of the ten main heat exchangers. This steam then flows to the main 46 steam header atop the 109 Building. Sufficient steam to drive the six drive turbines and to supply auxiliary areas uses flows to the turbine supply header the remainder is condensed in the dump condensers, completing the heat transfer to the circulating raw water system. Secondary coolant pressures, flow rates and Inventory are directly controlled coolant temperatures are dependent variables. 

Pb is proposed for use in an LFR at pressures close to 0.1 MPa. Pb has a higher melting point (327.5°C) and a significandy higher boiling point (1750°C) compared to that of Na, which significandy impacts the manner of operating a reactor. Also, it is a more inert liquid metal than Na. Due to that, the LFR has only two loops (1) a primary loop with Pb as a reactor coolant and (2) a secondary loop with water/steam as a steam Rankine power cycle. 

The TMI-2 reactor, the 880 MWe unit, was operating at 97% of rated power before the accident. Figure 2.4-2 is a simplified drawing that depicts the pre-accident conditions in the reactor coolant system. Figure 2.4-2 indicates a reactor coolant system pressure of 2150 psi (14.8 MPa), flow of subcooled water through both reactor coolant loops, a steam bubble in the pressurizer, and boiling of secondary water in... 

Measuring and recording/indicaling the reactor technical parameters such as temperature at various points in the reactor tank and at the heat exchanger lost water levels in the reactor and other tanks and in the reactor hall sump water flow-rates of the primary and secondary loops air flow-rates from the reactor tank space and in the reactor stack pressure at some points on the primary and secondary loops conductivity of reactor water before and after passing the water purification system of the primary loop. 

The PWR is more expensive to build because the reactor vessel must be stronger to withstand the higher water pressure, and there is a secondary coolant loop with pumps and so on. The BWR, while less expensive to build, is more complicated to service since the turbines are part of the primary coolant loop. The details of the core design are different as well. Approximately twice as many PWRs have been constructed as BWRs. 

A primary compressor increases the pressure of the entering ethylene gas (and propylene gas, which is added as a molecular weight control agent) from between 5 and 15 bar to about 250 bar. The secondary compressor further increases the gas pressure from 250 bar to the desired reactor pressure (approximately 2500 bar). An initiator is added to the gas as it enters the reactor. The reactor is operated to ensure a per-pass conversion of 15%-35% and is a wall-cooled reactor where the cooling water can be used to produce steam. The reaction mixture then enters the HP separator (-250 bar), where the mixture is flashed to produce two distinct phases a PE-rich melt phase and an ethylene-rich gas phase. The separated gas then enters the recycle loop. The ethylene gas is cooled before entering the secondary compressor. The PE enters the low-pressure separator. This low-pressure separator, also referred to as a hopper, performs the final degassing step. The separated ethylene gas is cooled and some components are removed. This step takes place... 

Gas cooled reactors use carbon dioxide under pressure as a recirculating heat transfer medium (coolant) between the hot nuclear reactor core and water in a secondary circuit in order to raise steam and electrical power in an otherwise conventional high pressure steam generator/turbine/condenser loop. The role played by ion exchange is denoted by systems A-D in Figure 8.22. 

Fast Breeder Test Reactor (FBTR) is a 40 MWt/ 13.2 MWe sodium cooled, mixed carbide fuelled, loop type reactor. It has two primary and secondary sodium loops and a common steam water circuit, which supplies high pressure, high temperature superheated steam to turbine generator (TG). Heat is rejected in cooling tower (Fig 1). A 100% capacity dump condenser is provided for reactor operation even when the TG is not in service. The mmn aim of the reactor is to generate experience in the design, construction and operation of sodium cooled fast reactors and to serve as an irradiation facility for the development of fuels and structural material for fast reactors. It achieved first criticality in Oct 85 with Mark I core... 

The second heat removal system is an independent cooling system (ICS), which includes, besides a part of primary and secondary circuit equipment, a loop separator-cooling condenser with natural circulation. Via this loop the heat is removed to the intermediate circuit water. This system ensures independent (from the turbine generator systems) reactor cooling and independent reactor plant operation at a constant power level up to 6 % N om at the nominal steam pressure. In case of total RI de-energizig the system ensures cooling of the reactor over several days. Connection/disconnection of ICS is realized with no operator action and without using external power supply systems. 

Nuclear heat transfer to desalination plant The nuclear power plant supplies steam to three turbogenerators and to a desalination plant (80 000 tonnes of desalinated water per day). To transfer the heat from the reactor to the intermediate heat exchangers six primary parallel sodium loops and six independent secondary sodium loops are provided. The turbine exhaust steam under the pressure of 0.6 MPa (6 bar) is supplied to the desalination plant, and the condensate produced with the temperature of about 100°C flows to a heater and a deaerator and is pumped by feedwater pumps to the natural circulation steam generators (Fig. 4). 

The core outlet temperature is set to 900°C, which is below what has been, or will be demonstrated, by the fuels in AVR (Arbeitsgemeinschaft Versuchs Reaktor), HTR-10, HTTR (High Temperature Test Reactor), etc. No safety hazard of water or steam ingress into the primary system exists, since all water and steam circulation is remotely located in the third loop. The nitrogen heater is essentially pressure balanced with a slightly higher secondary pressure to ensure that no fission products enter the secondary system in case of leaking tubes. 

The first experiment series included five Small Break LOCA tests with break in one hot leg of PACTEL. Three break sizes (2, 4 and 6 mm or 0.5, 2 and 4.4%, respectively) were used. Three tests included secondary system depressurization as an accident management measure. The operators of the loop also depressurized the secondary system by opening a steam generator relief valve. The core power in the tests was 80 kW corresponding to 1.8% of the scaled thermal power of the reference reactor. In all tests, only PSIS provided ECC water to the core. The initial temperature of the water in the CMT was 40°C. The primary pressure used in the tests was lower than the nominal operation pressure of PACTEL. Maximum operation pressure of the passive accumulators determined the upper limit of the experiment pressure (3.8 MPa). 

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