Designing a Core Lattice: Thermal-Hydraulics & Neutronics Considerations

In summary, when designing a core lattice, one has to consider thermal hydraulics and neutronics. Thermal hydraulics refers to the constraints imposed by the surrounding environment, while neutronics refers to the distribution of power within the reactor. There is an interplay between these two considerations, and one has to balance fuel rod diameter/cladding diameter/rod pitch while moderating to keep the reactor under control.
  • #1
candice_84
45
0
When designing a core lattice, which one of the thermal-hydraulics and neutronics consideration are important?
 
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  • #2
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
 
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  • #3
How about moderation-to-fuel ratio?
 
  • #4
candice_84 said:
How about moderation-to-fuel ratio?
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.
 
  • #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?
 
  • #6
candice_84 said:
Is it true,If we increase the fuel, the fuel utilization increases? therefore moderation-to-fuel ratio decreases and the reactor becomes under moderated?
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.
 
  • #7
Is there any advantage for the Russian design to be hexagonal over PWR lattice, or they just wanted to make it look different?
 
  • #8
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.
 
  • #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?
 
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  • #10
candice_84 said:
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?
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.
 
  • #11
candice_84 said:
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?
Neutron energy spectrum.
 
  • #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.
 
  • #13
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
 
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  • #14
Astronuc said:
Fast reactors generally use hexagonal lattices.[/url]

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?
 
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  • #15
candice_84 said:
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?
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.
 
  • #16
In the pdf file for LMFR it says that LMFR don't need emergency cooling systems but I don't understand it.
 
  • #17
candice_84 said:
In the pdf file for LMFR it says that LMFR don't need emergency cooling systems but I don't understand it.
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).
 
  • #18
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]
 
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  • #19
How important is it, to get an accurate burnup calculation and couple it with thermal-hyraulics?
 
  • #20
candice_84 said:
How important is it, to get an accurate burnup calculation and couple it with thermal-hyraulics?
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.
 
  • #21
Is it bad for the reactor to have high burnup?
 
  • #22
There are burnup limits on commercial fuel due to licensing constraints on the fuel behavior related to postulated accidents and activity release.
 
  • #23
What are the advantages and disadvantages of upward and Downward flow in PWR?
 
  • #24
candice_84 said:
What are the advantages and disadvantages of upward and Downward flow in PWR?
Is the question referring to core flow or upper head and by-pass flow?
 
  • #25
I mean core flow.
 
  • #26
candice_84 said:
I mean core flow.
OK. I thought so, but there is also a use of the terminology of downflow and upflow for the bypass region which relates to baffle-jetting in some PWRs. Usually downflow plants have been converted to upflow in the bypass.

OK - what can one say about flow of a liquid involved in heat transfer in terms of differential pressure (pressure drop), bouyancy, and coolant temperature?
 
  • #27
Well, I'd like to know about all of them: Pressure Drop, Buoyancy and Coolant Temperature. My own guess about pressure drop is that when the flow is downward inside the core, there would be less pressure drop therefore less pumping power which is an advantage of downward flow. But the actual flow in the pwr goes down in the downcomer and then flows upward which i don't think its a good design but there has to be a trade off that I am not aware of it.
 
  • #28
If one has a pipe and one measures the pressure at point A and down stream at point B, what can one say about the pressure at point A relative to point B? Assume that points A and B sit in the same horizontal plane with respect to the local gravitational field.

What would be the effect of point A being higher than point B, or point B being higher than point A?
 
  • #29
Pressure is higher at Point B, because P=[tex]\rho[/tex]gh. Since H is higher in point B, pressure is more at point B.
 
  • #30
candice_84 said:
Pressure is higher at Point B, because P=[tex]\rho[/tex]gh. Since H is higher in point B, pressure is more at point B.
What if A and B are at the same elevation (h), or Δh=0, and water is pump (forced convection) from point A to B?

If one has a closed loop of piping, including a pump, and the pump is pushing the water around the loop, where is the maximum pressure?
 
  • #31
I think pressure at the horizontal line is the same.
I cannot picture the second question, therefore I don't know where the pressure is high in the tube.
 
  • #32
candice_84 said:
I think pressure at the horizontal line is the same.
I cannot picture the second question, therefore I don't know where the pressure is high in the tube.
The primary cooling circuit or loop of a PWR is more or less a closed system, with the exeception of lines use to infuse or extract water for cooling pumps, boric acid, coolant sampling lines. The primary cooling system consists of the reactor core and pressure vessel (PV) internals, hot and cold legs, steam generators (headers and tubing), and cross over legs from the steam generator to reactor coolant pump on the cold leg. The pressurizer is attached to one of the hot legs and provides pressure to the system.

The highest pressure in the primary circuit is not the pressurizer, but at the outlet of the reactor coolant pump (RCP). The further one moves down stream from the RCP, the lower the pressure. The lowest pressure is at the inlet of the RCP. It's a bit like a voltage source in an electric circuit, the greatest electrical potential is at the + terminal, and the potential falls as one moves downstream around the circuit until returns to the - terminal (Kirchhoff's voltage law). The pressure drop across the RCP from inlet to outlet equals the pressure drop around the entire loop (continuity).

The pressure drop across the reactor core is about 25-26 psid. After the water enters the RPV through the nozzle with the cold leg, it then flows through the downcomer, through the lower plenum, up through the core support plate, through the core (flow channels between fuel rods), up through the upper core guide structure (where control rods are located) where it turns out to the hot nozzles.

Across the core, there is a pressure drop (pressure drops because of flow resistance). The temperature increases from the inlet (~280-292°C) to exit (~320-330°C). The reactor components in contact with the coolant are at those local temperatures. There may also be nucleate boiling in the hottest (combination of local coolant enthalpy/temperature and heat flux). The density of the coolant also changes.

The hot and cold nozzles are at about the same elevation, or the cold nozzles may be slightly lower than the hot nozzles. The coolant density is a function of temperature. The static head in the cold leg is of course greater than the static head in the reactor where there is a temperature gradient - and the hottest temperature is at the core exit.

So what are the advantages and disadvantages of downflow in the core vs the conventional upflow.

Think about component operating temperature, thermal hydraulics in the core, and safety issues.
 
  • #33
candice_84 said:
... But the actual flow in the pwr goes down in the downcomer and then flows upward which i don't think its a good design but there has to be a trade off that I am not aware of it.

Aside from any of the hydraulic effects Astronuc is discussing, think about what would happen if one of the main coolant pipes were to break. The vessel design with the downcomer allows the safety injection system to refill the vessel and flood the core. If you eliminate the downcomer, you would need to have the cold leg piping penetrate the vessel bottom - and a break in that cold leg piping would just drain away any safety injection flow. (And if you had flow 'up', it would be the hot leg piping that penetrates the lower head).

I think this is the main reason for the downcomer design (though there may be other reasons, like having the water provide shielding to the vessel material). Notice that the ECCS rules specify breaks in the main coolant loop piping - a break in the vessel itself is not a postulated design basis event.
 
  • #34
I didn't specifically call out the ECCS, but that comes under safety considerations.

One might be interested in 3.9.5 Reactor Pressure Vessel Internals in AP1000 DCD Rev. 17
http://adamswebsearch2.nrc.gov/idmws/doccontent.dll?library=PU_ADAMS^PBNTAD01&ID=083251077

The entire DCD can be found
http://adamswebsearch2.nrc.gov/idmws/ViewDocByAccession.asp?AccessionNumber=ML083230868

Chapter 4 contains the information on the reactor.
Chapter 3 Design of Structures, Components, Equipment and Systems
Chapter 5 Reactor Coolant System and Connected Systems

Earlier version of DCD is found at - http://www.nrc.gov/reactors/new-reactors/design-cert/ap1000.html#dcd
 
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  • #35
Can you guide me through calculating actual nuclear plants efficiency?
 
<h2>1. What is a core lattice in nuclear reactor design?</h2><p>A core lattice refers to the arrangement of fuel rods and control rods in a nuclear reactor. It is a critical component in the design of a nuclear reactor as it determines the thermal-hydraulic and neutronic properties of the reactor.</p><h2>2. How is the thermal-hydraulic behavior of a core lattice determined?</h2><p>The thermal-hydraulic behavior of a core lattice is determined through computer simulations and experimental testing. These methods help to analyze the flow of coolant through the lattice and its effect on the temperature distribution in the reactor.</p><h2>3. What are the key considerations in designing a core lattice?</h2><p>The key considerations in designing a core lattice are the thermal-hydraulic and neutronic properties. This includes factors such as fuel rod spacing, coolant flow rate, and control rod placement to ensure safe and efficient operation of the reactor.</p><h2>4. How does neutronics play a role in core lattice design?</h2><p>Neutronics refers to the study of the behavior and interactions of neutrons in a nuclear reactor. In core lattice design, neutronics is crucial in determining the distribution of neutron flux, which affects the efficiency and safety of the reactor.</p><h2>5. What are some challenges in designing a core lattice?</h2><p>Some challenges in designing a core lattice include achieving a balance between thermal-hydraulic and neutronic properties, optimizing fuel rod and control rod placement, and ensuring the structural integrity of the lattice under extreme conditions such as high temperatures and radiation levels.</p>

1. What is a core lattice in nuclear reactor design?

A core lattice refers to the arrangement of fuel rods and control rods in a nuclear reactor. It is a critical component in the design of a nuclear reactor as it determines the thermal-hydraulic and neutronic properties of the reactor.

2. How is the thermal-hydraulic behavior of a core lattice determined?

The thermal-hydraulic behavior of a core lattice is determined through computer simulations and experimental testing. These methods help to analyze the flow of coolant through the lattice and its effect on the temperature distribution in the reactor.

3. What are the key considerations in designing a core lattice?

The key considerations in designing a core lattice are the thermal-hydraulic and neutronic properties. This includes factors such as fuel rod spacing, coolant flow rate, and control rod placement to ensure safe and efficient operation of the reactor.

4. How does neutronics play a role in core lattice design?

Neutronics refers to the study of the behavior and interactions of neutrons in a nuclear reactor. In core lattice design, neutronics is crucial in determining the distribution of neutron flux, which affects the efficiency and safety of the reactor.

5. What are some challenges in designing a core lattice?

Some challenges in designing a core lattice include achieving a balance between thermal-hydraulic and neutronic properties, optimizing fuel rod and control rod placement, and ensuring the structural integrity of the lattice under extreme conditions such as high temperatures and radiation levels.

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