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Temperature questions

  1. Jan 5, 2009 #1

    Im wondering what kind of temperatures are typical for the primary and secondary coolant in a nuclear power plant. How much does the primary coolant temperature drop in the steam generator? I also need to know the pressure in the primary and secondary loop. Rough numbers are OK!

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  3. Jan 5, 2009 #2


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    High duty 4-loop 17x17 (193 assy) W-type NPPPs have a primary inlet temp of about 285-292°C and outlet temperature 315-330°C (nominal core pressure ~ 2250-2270 psia, pressurizer ~ 2235-2250 psia). Delta-T is ~ 35-36°C (66°F) with a little less across the vessel because of bypass flow. This is the same temperature drop across the primary side of the SG.

    I'll have to catch up later on the secondary side conditions, but IIRC, the pressure is about ~935-950 psia?
    Last edited: Jan 5, 2009
  4. Jan 5, 2009 #3


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    BWRs combine the reactor and steam generator.
    Rx Dome pressure is about 1040 psia.

    Feedwater inlet temperatures are between 360 to 420F.
  5. Jan 5, 2009 #4
    The other posters got their answers in while I was typing this (and answering a few phone calls). So, add it to what they said.

    The reactor vessel outlet temperature is between 600 and 620 F depending on the design and history. Most plants have reduced their temperature programs to limit temps to 600 (or not too much over that) due to stress corrosion problems.

    The vessel inlet (cold leg temp) is about 50 to 60 F less than the hot leg. For a 1100 MWe (3400 MWth) core the RCS flow is about 400,000 gpm (use that along with the pressure to calculate the enthalpy change, and then use the pressure and enthalpy to find the temperature at the core inlet). The RCS pressure is nominally 2250 psia.

    The secondary pressure is about 850 to 900 psia. Find the steam temp from the steam tables (it is saturated steam).

    Note that these are typical values, a given plant may operate with a different program. Some have Tcold constant, some have a decreasing Tcold (such that Taverage is constant).

    If you're not used to these units (F, psi, gpm) you'll have to convert them to SI yourself.
  6. Jan 5, 2009 #5
    Thanks for the quick response!
  7. Jan 5, 2009 #6


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    Here's a good ref on some secondary side characteristics

    Water Chemistry of Nuclear Reactor Systems Page 629 gives some characteristics of 3 US PWRs.

    Steam temp ~ 276-285°C with hot leg inlet 232-326°C. Some plants have done Thot reduction to prolong lifetimes of SG tubing (big problem with Inconel 600). Some plants have replaced SG tube bundles with Inconel 690 (favored in US), while the Europeans have favored Incoloy 800.

    I have one source that indicates secondary SG feedwater at ~220°C at about 6.45 MPa. The pressure seems a bit low to me. Another source indicates ~7 MPa FW pressure.
  8. Jan 5, 2009 #7
    Sequoyah- Westinghouse PWR primary 584F at 2100 psi appx.

    Sorry I don't know secondary.
  9. Jan 6, 2009 #8


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    Well the secondary steam, whether a PWR or BWR, must be at saturation temperature (unless it is a PWR with a counter-flow once-through steam generator ala TMI, in which case it would be a bit higher).
  10. Jan 7, 2009 #9
    Here's a comparison of the CANDU designs. The newest one, "Advanced CANDU" ACR-1000, has reactor coolant parameters of 12.6 MPa (A), 275 °C inlet; 11.2 MPa (A), 319 °C outlet; flow rate 14,560 kg/s (520 channels, each 28 kg/s max). The 4x steam generators' output parameters are 6.0 MPa (A), 276 °C. The reactor core nominally outputs 3,187 MW heat; the electric output of the power plant is ~1,200 MWe, depending (I think) on the cold water reservoir.

    Note that the Advanced CANDU primary coolant is in fact LIGHT water, not heavy water (deuterium oxide, D20) as in the other designs.

    http://www.aecl.ca/Reactors/ACR-1000.htm [Broken]

    http://www.aecl.ca/Assets/Publications/ACR1000-Tech-Summary.pdf [Broken]

    Here's a BWR (boiling water reactor) for contrast. The reactor coolant inlet is water but the output is steam!, This is GE's "Advanced BWR" ABWR, rated for 3,926 MW(th), 1,356 MW(e). Its reactor coolant parameters are 7.4 MPa (A), 278 C inlet; 7.2 MPa (A), 288 °C outlet; flow rate 14,500 kg/s. If I'm not mistaken, this is the only loop; I think BWRs have only one primary loop, which runs from the reactor to the turbine.



    Now here's a representative PWR (pressurized water reactor). This is Westinghouse' AP-1000, rated for 3,400 MW(th), 1,000 MW(e). Its reactor coolant parameters 16.0 MPa (A), 281 °C inlet; 15.6 MPa (A), 321 °C outlet; total flow rate 19,873 L/s (not sure how much mass this is). The two steam generators have inlet water at 227 °C, outlet steam at 5.8 MPa, 273 °C; their combined flow rate is 1,887 kg/s.


    http://www.nrc.gov/reactors/new-reactors/design-cert/ap1000/dcd/Tier 2/Chapter 5/5-1_r15.pdf

    http://www.nrc.gov/reactors/new-reactors/design-cert/ap1000/dcd/Tier 2/Chapter 10/10-1_r3.pdf

    There is a Gen IV concept using water at far high temperatures - beyond the critical point, where the liquid/gas phase change disappears. This would be above 22 MPa and 374 °C.


    At even higher temperatures, a gas coolant like helium can be used. This has been done at the THTR-300 thorium pebble bed reactor (commercial), and some other ones. I haven't tracked down THTR's coolant properties. But one example under development, the Pebble Bed Modular Reactor, is designed to operate between 9.0 MPa, 482 °C, and 8.7 MPa, 900 °C (reactor inlet/outlet). This is coupled to a Brayton cycle turbine. A Gen IV VHTR (Very High Temperature Reactor) wants to push this beyond 1000 °C, for thermochemical hydrogen production.




    I haven't found any information on the liquid sodium reactors - Monju, BN-600, EBR, and others. Nor the designs using molten salt coolant, or supercritical CO2.

    I've made some unit conversions here to improve clarity - gauge pressure to absolute pressure, psi to MPa, °F to °C, and so on. Errors are probably mine.
    Last edited by a moderator: May 3, 2017
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