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Reactor Design

  1. Aug 27, 2010 #1
    When designing a core lattice, which one of the thermal-hydraulics and neutronics consideration are important?
     
  2. jcsd
  3. Aug 27, 2010 #2

    Astronuc

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    Neutronics, or rather, power distribution goes hand in hand with thermal hydraulics.

    Basically one has to do the nuclear design subject to the constraint imposed by the thermal hydraulics (thermal boundary conditions).

    There's an interplay also with the moderation, moderator temperature coefficient, in addition to the critical heat flux (CHF) and the margin to CHF.

    One also has to be mindful of 10 CFR 50, App. A and the General Design Criteria, particularly GDC 10, 27 and 35, which are frequently cited in NUREG-0800, Standard Review Plan, particularly Section 4.2, Fuel System Design.

    http://www.nrc.gov/reading-rm/doc-collections/cfr/part050/part050-appa.html

    Reactor
    http://www.nrc.gov/reading-rm/doc-collections/nuregs/staff/sr0800/ch4/

    4.2 Fuel System Design
    4.3 Nuclear Design
    4.4 Thermal and Hydraulic Design
     
    Last edited by a moderator: Apr 25, 2017
  4. Aug 27, 2010 #3
    How about moderation-to-fuel ratio?
     
  5. Aug 27, 2010 #4

    Astronuc

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    That's one parameter.

    One has to do a balance between fuel pellet diameter/cladding diameter/ and rod pitch. One can open up the lattice which is cooler, but there is more leakage of neutrons.

    Or for a given pitch, one can make a smaller fuel rod diameter, but that drives of the heat flux for a given linear power.

    Of course, my example applies to a tubular design.

    One can make a core a bit more homegeous with spherical fuel particles which pretty much necessitates a gas like He, Ne, CO2 for heat transfer. Reactivity management can be more of challenge though.

    Then again, one could used a liquid fuel, e.g., molten salt, but that requires special considerations of the primary circuit and heat exchangers.
     
  6. Aug 27, 2010 #5
    Is it true,If we increase the fuel, the fuel utilization increases? therefore moderation-to-fuel ratio decreases and the reactor becomes under moderated?
     
  7. Aug 27, 2010 #6

    Astronuc

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    Yes, and that would harden the spectrum. In BWRs, one can reduce flow and increase voiding in an assembly and harden the spectrum. The harder spectrum produces more Pu-239 from conversion of U-238, which can increase fuel utilization. That operation is called spectral shift.

    So-called standard 17x17 PWR fuel has a UO2 pellet diameter of 0.3225 inch (8.19 mm), and cladding OD of 0.374 inch (9.5 mm). The lattice pitch is 0.496 inch (12.6 mm).

    In the 1980s, Westinghouse introduced an optimized 17x17 which used a slightly thinner fuel rod which used a pellet with an OD of 0.3088 inch (7.84 mm) and cladding OD of 0.360 inch (9.14 mm) in the same lattice. The same fuel rod dimension is used VVER-1000 fuel assembly, but the lattice is triangular/hexagonal.

    With the fatter fuel rod, one can load more fuel into the core, which can be used to reduce enrichment (and SWU costs), or go for slightly longer cycle length/exposure.
     
  8. Aug 28, 2010 #7
    Is there any advantage for the Russian design to be hexagonal over PWR lattice, or they just wanted to make it look different?
     
  9. Aug 28, 2010 #8

    Astronuc

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    With respect to advantage, one has to look at the fuel moderator ratio, and critical heat flux.

    The pitch of the VVER-1000 fuel rod is 0.5 inch (12.7 mm), but that is on a triangular geometry rather than a square lattice geometry. One should also look at the hydraulic diameter, as well as the fast to thermal flux ratio, or relative proportions in different neutron energy groups.

    I understand that the VVER lattice is well suited for thorium based plants.
     
  10. Aug 28, 2010 #9
    Where do they fit a flux detector in either pwr or vver? Is it inside the fuel rod? also does it conflict with absorber rods? Also when you mentioned about spectrum what spectrum exactly you are pointing on?
     
    Last edited: Aug 28, 2010
  11. Aug 28, 2010 #10

    Astronuc

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    One can do both in-core and ex-core detectors. The problem with ex-core detectors is that the neutron flux is very low at the core periphery.

    In-core detectors are usually placed in an instrument tube, and the incore instrumentation includes thermocouples for core inlet or core outlet. The instrument tube is like a guide tube, but it is usually centrally located in an assembly of odd-numbered rows (15 x15 or 17x17) or slightly off-set from center in 14x14 or 16x16 assemblies, except for CE plants which have 4/5 guide tubes displacing 4 fuel rods.

    AREVA's EPR is designed to use a guide tube instead of an instrument tube, and the EPR fuel assembly has 265 fuel rods instead of the traditional 264 fuel rods of a 17x17 lattice. W's AP-1000 and Mitsubishi's US-APWR use the traditional centrally located instrument tube. The instrumentation is placed in an unrodded location, i.e., no control rods in the assembly. AREVA also uses a pneumatic activation analysis system, which is based on Siemens technolgoy, in addition to typical SPN detector strings.

    Some plant designs had incore probes inserted from below the core, but the modern plants favor insertion from the top.
     
  12. Aug 28, 2010 #11

    Astronuc

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    Neutron energy spectrum.
     
  13. Aug 28, 2010 #12
    Can we use the same 17x17 fuel assembly in fast reactors? also what materials are suited as an absorber for fast reactor, because I was checking the cadmium's cross section and noticed it doesn't absorb at high energy level.
     
  14. Aug 28, 2010 #13

    Astronuc

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    Fast reactors generally use hexagonal lattices.

    The reference structural material for fast reactors is SS316 - both core and fuel structures. Other steels have been developed for improved creep resistance and reduced swelling.

    One can find information here on the Clinch River Breeder Reactor
    http://www.osti.gov/bridge/servlets/purl/4281663-9ySuUJ/4281663.pdf

    Otherwise there is EBR-I, II at INL (INEL).

    One can search on EBR-II and FFTF driver fuel.

    Unfortunately, when FFTF was mothballed and trashed, a lot of fuel material was dumped and a lot of information was tossed. :(

    A good reference is the Proceedings from an ASTM conference: Effects of radiation on materials / Radiation-induced changes in microstructure.
    Example: http://books.google.com/books?id=k5j2P1xKKqEC&pg=PA146&lpg=PA146
     
    Last edited: Aug 28, 2010
  15. Aug 28, 2010 #14
    Since there is no moderator involved in fast reactors, I assume hexagonal lattices perform better thermal-hydraulics than rectangular 17x17 in general. Is this a right assumption? Can you provide me a reference that shows how to mathematically do thermal-hydraulics analysis?
     
    Last edited: Aug 28, 2010
  16. Aug 28, 2010 #15

    Astronuc

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    I believe that the tringular/hexagonal lattice is better from a neutronic standpoint. The neutron leakage is more before the mean free path of a fast neutron is much greater in a fast reactor than in a thermal reactor. In the latter, fast neutrons slow down relatively quickly.

    The thermalhydraulics in a fast reactor is a bit different since it is liquid metal, Na or NaK, or if one wants to get really exotic, Li. One has to be concerned with the positive void coefficient.

    I believe that the control elements were B4C based. It's been a while since if reviewed the technology.

    I'll see if I can find a good reference on fast reactor T/H. Probably Waltar and Reynolds is the best text on fast reactors.
     
  17. Aug 28, 2010 #16
    In the pdf file for LMFR it says that LMFR don't need emergency cooling systems but I don't understand it.
     
  18. Aug 28, 2010 #17

    Astronuc

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    Probaby there is a decay heat removal system, but not an ECCS like that of an LWR. The liquid metal coolant has excellent thermal conductivity, and I suspect that in shutdown, there is an effective heat removal system and one does not have to be concerned of the core overheating or the coolant boiling as the case with an LWR. The problem for LWRs is that they have to operate under high pressure to maintain a liquid phase. Na systems operate at a few atmospheres. IIRC, a Na loop at 0.20 or 0.5 MPa (about 2 to 5 atm) provides much the same heat transfer coefficient as pressurized water at 2250 psia (15.5 MPa), 290 C (563 K).
     
  19. Aug 29, 2010 #18

    Astronuc

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    This is a reasonable good reference for LWR fuel design data - http://www.neimagazine.com/journals/Power/NEI/September_2004/attachments/NEISept04p26-35.pdf [Broken]
     
    Last edited by a moderator: May 4, 2017
  20. Aug 30, 2010 #19
    How important is it, to get an accurate burnup calculation and couple it with thermal-hyraulics?
     
  21. Aug 30, 2010 #20

    Astronuc

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    It depends on the calculation.

    For economics, it's important with respect to determining utilization.

    For fuel performance, it's important to get local (nodal) burnup as well as radial distribution of burnup for accurately predicting PCI and transient behavior. Many of the materials properties, e.g., UO2 thermal conductivity and swelling are strongly dependent on local burnup, while cladding mechanical properties are fluence and temperature dependent, and cladding oxidation is time and temperature, and to some extent water chemistry, dependent.

    There are certain regulatory and licensing requirements that are dependent on burnup.

    It is primarily fuel performance and certain operational/safety limits (and margins) for which accurate modeling of thermal-hydraulics is important. Properties dependent on the coolant local coolant conditions are important.
     
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