DOI: 10.1615/AtoZ.s.steam_generators_nuclear


The NSSS (Nuclear Steam Supply System) is a relatively recent development, and has been in use for about thirty years. During this time, there were constructed and put in operation 298 Pressurized Water Reactors (PWR), 81 of which are in the U.S.; 100 Boiling Water Reactors (BWR), 38 of which are in the U.S.; 19 light-water cooled graphite-moderated reactors (LGR) and 50 pressurized heavy water moderated and cooled reactors (PHWR)—all over 30 MW. In addition, there were expected to be in operation 163 more PWRs, 56 BWRs, 12 LGRs and 18 PHWRs. Here the attention is focused only on the nuclear steam generators of a PWR system, which is shown schematically in Figure 1.

Heat, which is produced in the core inside the pressure vessel, is converted by the primary fluid, which is pumped through the pressure vessel, from the core to the system generator. In the steam generator the primary fluid, water at 150 bar, exchanges heat with the secondary fluid, water at 75 bar, and causes it to boil. The steam, from the steam generator, passes through the turbine, condenser, and is pumped back into the steam generator (SG) as feed water.

There are two types of SG: the U-tube SG and the once through SG, as shown in Figures 2 and 3.

U-Tube Steam Generators

In Figure 2 is depicted a U-tube steam generator, which is installed in a wide variety of commercial NSSS, from the early (1957) 90-MWe Shipping port 4 loops system to new 1300-MWe plants. Two typical systems are characterized in the following:

Rated power, MWe8221,300
Rated heat output, MWth2,4413,819
System pressure, bar153153
Primary coolant flow rate, 104 kg/hr45.367.9
Primary coolant temperature, °C 318,7 332.0
Number of loops34
Number of pumps34
Steam generators  
   Shell side design pressure, bar53.477.0
   Steam flow at full load, 106 kg/hr4.87.9
   Steam temperature, °C268.9294.1
   Feedwater temperature, °C225.4239.4
   Number of tubes3,3886,970
   Diameter of tubes, mm22.217.4
   Heat transfer area, m24,784 7,665

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Figure 1. Schematic view of a PWR power plant.

U-tube steam generator.

Figure 2a. U-tube steam generator.

Schematic diagram of a U-tube SG.

Figure 2b. Schematic diagram of a U-tube SG.

Reactor coolant enters the steam generator at the inlet nozzle (on the bottom left-hand side of Figure 2) and enters the tube bundle in the hot leg, completing the U-bend through the cold leg to the primary outlet.

Feedwater enters the side of the shell. It flows down through an annulus just inside the shell (downcomer), where it mixes with water coming from the separator deck. The water enters the heating surface (tube bundle) at the bottom and is increasing in quality as it rises through the steam generator. The steam-water mixture enters the steam water separators—where steam is passed through to the driers and then to the steam nozzle; the water is recirculated, through the downcomer, to the bottom of the heating surface.

The shell, outside of the tubes and the tubesheet form the steam production boundaries. Within the shell, the tube bundle is surrounded by a wrapper (or shroud). The tube material is usually Inconel Alloy 600 (though it would have been better to have thermally treated Inconel 690). Tube support plates, with quatrofoil holes or egg crate supports, hold the tubes in uniform pattern along their length. The U-tube region of the tube is additionally supported by antivibration bars. Vents, drains, instrumentation nozzles, and inspection openings are provided on the shell side of the units.

The UTSGs are characterized by a widening of the shell about 1/3 of the height of the steam generator. This is done in order to increase the area available for the separators at the separator deck.

In the UTSG there is always a water level in the downcomer, in order to balance the pressure losses of steam-water mixture as it flows through the tube bundle. The pressure drop in the tube bundle is due to friction along the tubes, pressure drop at the tube support plates and separators, acceleration of the flow, and hydrostatic head. Quality in the downcomer (e.g. steam bubbles) reduces the density, and thus the available head, substantially. This can be a result of imperfect separation of steam water at the separators and is termed carryunder.

The ratio of flow rate of the steam water mixture which flows through the SG tube bundle, to the flow rate of steam out of the steam nozzle, is called the circulation ratio. It is desirable to maintain a high circulation ratio (may be over 5) to reduce concentration of chemicals, debris, etc., in various places in the SG. Current SGs have frequently circulation ratios of around 3, which is undoubtedly one of the causes of the recently mounting difficulties with these units.

The design of SGs is a complicated procedure, which involves many steps, iterations and interaction with other components of the system.

The NSSS, as part of the power station, is designed to minimize the power cost. This consideration is subject to many constraints — the primary one being safety. Other constraints are imposed by availability of major equipment (e.g., primary coolant pumps), manufacturing capabilities (e.g., can vessel be fabricated), shipping consideration, etc.

Thus, the design of the steam generators is subject to many outside constraints and requirements. For example, the primary fluid condition (i.e., temperatures, allowable pressure drop in the steam generator) is determined mostly by the reactor design and availability of pumps. The performance requirements, i.e., steam pressure, temperature, and flow rates, are determined mostly by the turbine design and are part of the system performance. It follows, therefore, that the tube bundle size (namely the required heat transfer area, as well as allowable pressure drop of the primary fluid) is determined by system considerations.

The preliminary structural design is performed in accordance with ASME's boilers and Pressure Vessel Code, Section III. (See Pressure Vessels.) Also, steam generators, as all power plant components, are required to be designed to withstand various accident situations. See, for example, Blowdown. For steam generators, this consists of the following conditions:

  • Small steam line break, loss of feedwater, turbine trip, etc.

  • LOCA (Loss of Coolant Accident)

  • MSLB (Main Steam Line Break)

  • SSE (Safe Shutdown Earthquake)

The detailed design includes detailed structural and thermal-hydraulic analyses and studies, and is later used to prove to the customer and regulatory agencies that the steam generator design is in compliance with ASME code, NRC regulations, etc.

Once Through Steam Generators

The OTSG (Figure 3) is typically associated with a NSSS which has the following general characteristics:

Rated power, MWe8601,300
Rated heat output, MWth2,5683,760
System pressure, bar149153
Primary coolant flow rate, 106 kg/hr29.535.6
Primary coolant temperature °C317.8329.7
Number of loops22
Number of pumps44
Number of steam generators71,484.0
Steam flow at full load, 106 kg/hr299308
Steam superheat at full load, °C19.419.4
Feedwater temperature, °C235241
Number of tubes15,53016,000
Diameter of tubes, mm15.915.9
Once through SG.

Figure 3a. Once through SG.

Schematic diagram of a once through SG.

Figure 3b. Schematic diagram of a once through SG.

Reactor coolant water enters the steam generator at the top, flows downward through the tubes and out at the bottom. The high pressure parts of the unit are the hemisphere heads, the tubesheets and the straight tubes between the tubesheets. The tube material is Inconel Alloy 600. Tube support plates, with trefoil holes, hold the tubes in a uniform pattern along their length. The unit is supported by a skirt attached to the bottom tubesheet.

Figure 3b indicates the flow paths on the steam side of the unit. Feedwater enters the side of the shell. It flows down through an annulus just inside the shell where it is brought to saturation temperature by mixing it with steam. The saturated water enters the heating surface at the bottom and is converted to steam and superheated in a single pass upward through the generator.

The shell, outside of the tubes, and the tubesheets form the boundaries of the steam producing section of the vessel. Within the shell, the tube bundle is surrounded by a shroud, which is in two sections, with the upper section the larger of the two in diameter. The upper part of the annulus between the shell and the baffle is the superheater outlet, while the lower part is the feedwater inlet heating zone. Vents, drains, instrumentation nozzles and inspection openings are provided on the shell side of the units.

Superheated steam is produced at a constant pressure over the power range. At full power, the steam temperature of 300°C provides about 19°C of superheat. As load is reduced, steam temperature approaches the reactor outlet temperature, thus increasing the superheat slightly. Below 15% load, steam temperature decreases to saturation.

Recent Problems with SGs

The NSSS provided in the past service which was mostly safe and trouble free. In the last 20 years, however, there is an increasing number of PWR nuclear steam generators which develop technical problems, such as denting, intergranular attack (IGA), vibration of tubes which cause wear and fatigue, wastage of tubes, pitting, erosion-corrosion, water hammer, etc. Any of these can lead to a breach of the integrity of the tubes and to leakage of primary (contaminated) coolant into the secondary fluid. Since the secondary fluid is leaving the containment vessel to the turbines, it must not be radioactive, and must not be contaminated by primary fluid. Therefore, when leaking tubes are detected, the plant must be shut down for repairs, and replacement of SGs, at great costs and loss of revenue.


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Green, S. J. and Hetsroni, G. (1995) PWR Steam Generators, Int. J. Multiphase Flow, 21 (Annual Review). DOI: 10.1016/0301-9322(95)00016-Q

Hetsroni, G. (1982) Handbook of Multiphase Systems, Hemisphere-McGraw-Hill.

Steam Generation Reference Book, EPRI, 1985.

Singhal, A. K. and Srikkantiah, G. (1991) A Review of Thermal Hydraulic Analysis Methodology for PWR SG and ATHOS 3 code Application, Progress in Nuclear Energy, 25, 7-70. DOI: 10.1016/0149-1970(91)90041-M

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