**4. Helical-coil steam generators**

#### **4.1. Description and main characteristics**

Many future nuclear reactor projects, especially innovative small- and medium-sized reactor systems [12], are expected to use helically coiled pipes for the steam generators. Their favourable characteristics justify the helical tube SG development in the nuclear field. In particular, helical tubes provide enhanced heat and mass transfer rates, a higher critical heat flux during boiling and evaporation and a better capability to accommodate the thermal expansion, in addition to allow a more compact design of the SG. Helical coils are particularly attractive for small and medium modular reactors (SMRs) since many of them adopt an integral layout. Compactness and efficiency improvement become particularly important for this type of reactors, as all the primary system components are located inside the reactor vessel.

The helical coil SG design and operation will be explained on the example of the IRIS reactor [13]. The international reactor innovative and secure (IRIS), an integral, modular, mediumsized (335 MWe) PWR, has been under development since the turn of the century by an international consortium led by Westinghouse and including over 20 organizations from nine countries. IRIS features an integral vessel that contains all the major reactor coolant system components, including the reactor core, pumps, the steam generators and the pressurizer. The unique IRIS safety-by-design approach provides a very powerful first level of defence in depth approach by eliminating accidents, at the design stage, or decreasing their consequences and probabilities when outright elimination is not possible. There are no primary system pipings, and a large-break loss-of-coolant accident (LOCA), related to a double-ended break of a primary leg pipe, is avoided. The passive safety systems increase plant safety by providing core decay heat removal during accident conditions even where there is no electrical power supply.

Steam generators are located in the space between the core barrel, precisely the shroud enclosing the riser section, and the reactor vessel wall. There are eight SG units in total, designed as once-through heat exchanging units (**Figure 12**). They are made of helical tubes with secondary fluid flowing inside the tubes. The feedwater header is located at the bottom of the SG module, while the steam header is positioned at the top of the steam generator. The tubes are wrapped around the inner supporting column. The primary cooling water flows outside of the tubes, through the tube bundle. The primary reactor coolant pumps are installed above the steam generators and drive coolant from the top to the bottom of the SG. Thus, a countercurrent flow regime is developed inside the SG, the primary coolant flows from the top to the bottom of the SG and the secondary fluid flows in the opposite direction.

**Figure 12.** Simplified IRIS helical-coil SG flow paths.

Operator action to rapidly depressurize secondary side to 2 MPa using SG relief valves led to fast primary-side cooldown and depressurization. The primary and secondary systems were closely coupled with little difference in pressures and temperatures. Both systems were in saturated conditions, temperature depending on the saturation pressure. Since the core decay heat was only a couple of percents of the total power and the heat transfer area in the SG was very large, the temperature difference was only few kelvins, and, thus, pressures were almost the same. With the operator action, the CST was emptied at ~202,500 s. That was 30,000 s earlier than in the case without any operator actions due to higher AFW flow required during SG steam extraction to satisfy prescribed RCS cooldown rate. Reducing the primary pressure to 2 MPa enabled the actuation of accumulators' water injection which quenched the core. Therefore, although secondary-side heat sink was lost earlier, the larger RCS water inventory increased the time to core damage. In the case with no AFW flow, the core heat-up started after 2.2 h. In the case with the AFW availability, the core temperatures started to increase after 65 h for the case without SG depressurization, and after 70 h when the operator controlled the SG pressure.

Many future nuclear reactor projects, especially innovative small- and medium-sized reactor systems [12], are expected to use helically coiled pipes for the steam generators. Their favourable characteristics justify the helical tube SG development in the nuclear field. In particular, helical tubes provide enhanced heat and mass transfer rates, a higher critical heat flux during boiling and evaporation and a better capability to accommodate the thermal expansion, in addition to allow a more compact design of the SG. Helical coils are particularly attractive for small and medium modular reactors (SMRs) since many of them adopt an integral layout. Compactness and efficiency improvement become particularly important for this type of reac-

The helical coil SG design and operation will be explained on the example of the IRIS reactor [13]. The international reactor innovative and secure (IRIS), an integral, modular, mediumsized (335 MWe) PWR, has been under development since the turn of the century by an international consortium led by Westinghouse and including over 20 organizations from nine countries. IRIS features an integral vessel that contains all the major reactor coolant system components, including the reactor core, pumps, the steam generators and the pressurizer. The unique IRIS safety-by-design approach provides a very powerful first level of defence in depth approach by eliminating accidents, at the design stage, or decreasing their consequences and probabilities when outright elimination is not possible. There are no primary system pipings, and a large-break loss-of-coolant accident (LOCA), related to a double-ended break of a primary leg pipe, is avoided. The passive safety systems increase plant safety by providing core decay heat removal during accident conditions even where there is no electrical power supply. Steam generators are located in the space between the core barrel, precisely the shroud enclosing the riser section, and the reactor vessel wall. There are eight SG units in total, designed as once-through heat exchanging units (**Figure 12**). They are made of helical tubes with

tors, as all the primary system components are located inside the reactor vessel.

**4. Helical-coil steam generators**

182 Heat Exchangers– Advanced Features and Applications

**4.1. Description and main characteristics**

The tubes are set up in annular rows and connected to steam and feedwater headers (**Figure 13**), which, on the other hand, are connected to feedwater and steam lines piping by nozzles mounted on the external surface of the reactor vessel wall. At the tube inlet, orifices which reduce fluid flow are installed in order to maintain uniform flow distribution across the tubes and to prevent parallel channel flow instabilities. The pressure drops at these orifices are of the same order of magnitude as the pressure drops in the tubes. The main IRIS reactor helical-coil SG characteristics and parameters are shown in **Table 3**.

**Figure 13.** IRIS SG steam header (taken from Ref. [13]).


**Table 3.** IRIS helical-coil SG dimensions and operating parameters.

The tubes are set up in annular rows and connected to steam and feedwater headers (**Figure 13**), which, on the other hand, are connected to feedwater and steam lines piping by nozzles mounted on the external surface of the reactor vessel wall. At the tube inlet, orifices which reduce fluid flow are installed in order to maintain uniform flow distribution across the tubes and to prevent parallel channel flow instabilities. The pressure drops at these orifices are of the same order of magnitude as the pressure drops in the tubes. The main IRIS reactor

helical-coil SG characteristics and parameters are shown in **Table 3**.

184 Heat Exchangers– Advanced Features and Applications

**Figure 13.** IRIS SG steam header (taken from Ref. [13]).

Primary and secondary fluid temperature profiles are shown in **Figure 14**. The primary fluid is water, and its temperature continuously increases along the height of the SG (decreases in the flow direction). On the secondary side, subcooled water enters the tubes, heats up, boils and produces superheated steam in the upper part of the tubes. During boiling, steam temperature slightly decreases due to a small pressure drop across the tubes. The total pressure drop on the secondary side, from the feedwater inlet to the steam outlet, is 323 kPa and in the boiling part of the tubes 140 kPa. The void fraction at point when superheating starts is 0.98.

#### **4.2. Computational model and numerical simulation of the IRIS reactor**

The RELAP5 code was used to model the facility and simulate accident conditions. Although the code lacks appropriate correlations for the heat transfer coefficient in helical pipes, comparison with detailed calculation [14] shows a good agreement between the results. The worldwide experience and quality models inside the RELAP5 were the reasons for selecting it to perform IRIS preliminary safety assessment studies [15].

The nodalization of the IRIS nuclear power plant is shown in **Figure 15**. On the left side of the figure, the reactor core is in the lower part (CV 110), the riser in the middle part (CVs 120–123) and pressurizer in the upper part (CV 130). On the right side, the pump (CV 191) is below the pressurizer and connected to the steam generator (CV 211). The SG annular dead space surrounding the SG shell and the inner inactive region are modelled as CV 240 and CV 241, respectively. The lower downcomer region (CV 101) is below the steam generator. The secondary side of the steam generator is modelled with CV 271. All eight steam generators are modelled, but only one is presented in the figure. The primary side of the SG was represented with 25 control volumes and the secondary side with 50 control volumes. That fine nodalization was obtained during the optimization process that resulted with stable and correct steady-state performance. Parallel channel flow instabilities were eliminated by modifying pressure loss coefficients. The heat transfer coefficient was also adjusted using multiplication factor to accommodate the inability of the RELAP5 to model curved helical-coil geometry. In that geometry eddy flow currents are created inside the coils, which promote mixing of the fluid and, thus, increase heat transfer capability at the expense of higher pressure drops [16].

**Figure 14.** Primary and secondary fluid temperature profiles in the IRIS NPP.

Operation and Performance Analysis of Steam Generators in Nuclear Power Plants http://dx.doi.org/10.5772/66962 187

120–123) and pressurizer in the upper part (CV 130). On the right side, the pump (CV 191) is below the pressurizer and connected to the steam generator (CV 211). The SG annular dead space surrounding the SG shell and the inner inactive region are modelled as CV 240 and CV 241, respectively. The lower downcomer region (CV 101) is below the steam generator. The secondary side of the steam generator is modelled with CV 271. All eight steam generators are modelled, but only one is presented in the figure. The primary side of the SG was represented with 25 control volumes and the secondary side with 50 control volumes. That fine nodalization was obtained during the optimization process that resulted with stable and correct steady-state performance. Parallel channel flow instabilities were eliminated by modifying pressure loss coefficients. The heat transfer coefficient was also adjusted using multiplication factor to accommodate the inability of the RELAP5 to model curved helical-coil geometry. In that geometry eddy flow currents are created inside the coils, which promote mixing of the fluid and, thus, increase heat transfer capability at the

expense of higher pressure drops [16].

186 Heat Exchangers– Advanced Features and Applications

**Figure 14.** Primary and secondary fluid temperature profiles in the IRIS NPP.

A small-break LOCA was analyzed to demonstrate steam generator behaviour in accident conditions. The initiating event was the rupture of a chemical and volume control system pipe connected to the upper annular pump suction plenum of the reactor vessel. IRIS is designed to limit the loss of coolant from the vessel rather than relying on systems to inject water into the reactor vessel. It has a compact, small diameter, high-design pressure containment that assists in limiting the blowdown from the reactor vessel by providing a higher back pressure in the initial stage of the accident. Furthermore, four trains of passive emergency heat removal systems (EHRS) help to depressurize the primary system by condensing steam, coming out of the reactor core, on the steam generators tubes (depressurization without the loss of mass), and to remove the decay heat. Finally, it features automatic depressurization system to condense steam, released from the top of the pressurizer, in the pressure suppression pool located inside the containment.

Following the initiating event, the LOCA mitigation signal is rapidly actuated, the reactor and reactor coolant pumps are tripped and the four EHRS subsystems are actuated by closing the main feed and steam isolation valves and by opening the fail-open valves in the EHRS return lines connected to the SG feed lines. The EHRS is composed of pipes, valves and heat exchangers submerged in the water tank outside the containment. After the initial decrease of the coolant inventory in steam generators caused by the isolation of the feedwater flow, the EHRS operation restores the SG secondary fluid mass (**Figure 16**) and enables heat removal out of the steam generators (**Figure 17**). It does not take long for the situation to stabilize to ensure safe reactor conditions. Equalization of reactor vessel and containment pressures marks the end of the blowdown phase and start of a long-term cooling phase by the continued operation of the EHRS, with the pressure being slowly reduced as the core decay heat decreases.

**Figure 16.** SG secondary-side fluid mass in IRIS NPP.

**Figure 17.** Heat transfer in IRIS NPP steam generators.
