Thermal hydraulic design of nuclear reactor core

[Nucengable]

By a simple procedure , what should I do when I'm going through the thermal hydraulic design of nuclear reactor core...?
..
I put initial guesses for the core dimensions ( fuel , clad , gap , length)
initial guess for the fuel element pitch
desired power ...
I've found q'' critical heat flux based on the correlation in "Nuclear Systems I - Thermal Hydraulic Fundamentals - Todreas"
I've found q'max ( based o DNB =1.3 )
q'avg based on ( hot spot factor =2 )
and based on q'max I've found the temperature distribution in one single fuel rod
...
The results
after assuming
we have only 1 assembly inside the core that produces 17.6 MWth
All fuel rods have the same enrichment
the reflector will take care of the leakage

to produce the desired power we need
188 fuel rods inside the assembly
Rf = 0.405 cm (fuel pellet radius)
Tg = 0.005 cm (gap thickness)
Tc = 0.05 cm (clad thickness)
Lr = 360 cm (fuel rod length)
fuel element pitch (p) = 1.3 cm
Pressure 15.5 Mpa

with coolant flux flow rate 436 kh/hr.cm2 in one single channel
I've found that
ΔTF ( temperature drop through the fuel pellet) = 1656.5 C
ΔTg ( temperature drop through the gap) = 511.3 C
ΔTC ( temperature drop through the clad) = 95 C
ΔTC ( temperature drop through the coolant) = 18 C
..
does this sounds correct because I sill don't have that sense of the numbers...!

 

[Chronos]

I believe Westinghouse has some data you may find useful.

 

[Astronuc]

By a simple procedure , what should I do when I'm going through the thermal hydraulic design of nuclear reactor core...?
..
I put initial guesses for the core dimensions ( fuel , clad , gap , length)
initial guess for the fuel element pitch
desired power ...
I've found q'' critical heat flux based on the correlation in "Nuclear Systems I - Thermal Hydraulic Fundamentals - Todreas"
I've found q'max ( based o DNB =1.3 )
q'avg based on ( hot spot factor =2 )
and based on q'max I've found the temperature distribution in one single fuel rod
...
The results
after assuming
we have only 1 assembly inside the core that produces 17.6 MWth
All fuel rods have the same enrichment
the reflector will take care of the leakage

to produce the desired power we need
188 fuel rods inside the assembly
Rf = 0.405 cm (fuel pellet radius)
Tg = 0.005 cm (gap thickness)
Tc = 0.05 cm (clad thickness)
Lr = 360 cm (fuel rod length)
fuel element pitch (p) = 1.3 cm
Pressure 15.5 Mpa

with coolant flux flow rate 436 kh/hr.cm2 in one single channel
I've found that
ΔTF ( temperature drop through the fuel pellet) = 1656.5 C
ΔTg ( temperature drop through the gap) = 511.3 C
ΔTC ( temperature drop through the clad) = 95 C
ΔTC ( temperature drop through the coolant) = 18 C
..
does this sounds correct because I sill don't have that sense of the numbers...!

188 fuel rods per assembly is close to the number for a 14x14 assembly, and a typical power level would be 13.6 MW/assembly for a 14x14 assembly, 14.5 to 25.7 MW/assembly for a 15x15 assembly, ~16.5 MW/assembly for a 16x16 assembly, and 17.7 to 18.6 MW/assembly for a 17x17 assembly. The smaller the diameter, the more limiting the heat flux for a given set of coolant flow rate, saturation temperature, coolant pressure and core hieght.

Some published fuel design data are found in: http://www.neimagazine.com/journals/Power/NEI/September_2004/attachments/NEISept04p26-35.pdf



Some typical PWR fuel design numbers are:

Lattice    LP   FR   GT  IT   FOD    GAP    CID    COD   PITCH  

 14x14   196  176  20   1  0.928  0.020  0.948  1.072  1.43
 15x15   225  204  20   1  0.928  0.020  0.948  1.072  1.43
 16x16   256  232  24   1  0.911  0.019  0.93   1.075  1.43
 17x17   289  264  24   1  0.819  0.017  0.836  0.95   1.263
 18x18   324  300  24   1  0.805  0.017  0.822  0.95   1.263

LP = number of lattice positions
FR = number of fuel rods
GT = number of guide tubes
IT = number of instrument tubes (optional) - one guide tube may serve as IT

FOD = fuel pellet outer diameter (OD)
GAP = diamteral fuel cladding gap
CID = cladding inner diameter (ID)
COD = cladding outer diameter (OD)
PITCH = distance between centers of adjacent fuel rods. (need to verify for 16x16)

For fuel rods, there is considerable variation for diametral/radial dimensions within the fuel rod, i.e., dimensions within the cladding envelope.

Guide Tube OD ~ Pitch - spacer grid strip thickness

Typical core height ~ 3.66 m (12 ft) or 4.27 m (14 ft)

Plenum height ~ 18-20 cm for 3.66 m core

See - EUR 20056 EN, Main Characteristics of Nuclear Power Plants in European Union and Candidate Countries
ec.europa.eu/energy/nuclear/studies/doc/other/eur20056.pdf
A1.3.6 Technical Data for REP900, page 84 of 163

For REP 900 (3 coolant loops, 157 assembies/core)

Core inlet temp         286°C
Core outlet temp       323°C
Tsat                       345°C
Psat                       155 bar
Core inlet pressure    157-158 bar
Coolant mass flux      ~3500-3600 kg/m2-s
Pressure drop - core  ~1.9 bar

Typical fuel centerline temperature 900-1400°C

 

[Nucengable]

Thank you very much

 

[alphy]

I cannot confirm if your calculations are correct without reviewing the methodology and assumptions used. However, it is important to follow a systematic approach when designing a thermal hydraulic system for a nuclear reactor core.

Firstly, it is crucial to understand the requirements and constraints of the system, such as the desired power output and the available space for the reactor core. This will help in determining the initial guesses for the core dimensions and fuel element pitch.

Next, it is important to consider the critical heat flux and maximum heat flux values in order to ensure safe operation of the reactor. These values can be determined using established correlations, as mentioned in "Nuclear Systems I - Thermal Hydraulic Fundamentals - Todreas".

Furthermore, the temperature distribution in the fuel rods must be calculated to ensure that the desired power output can be achieved without exceeding the temperature limits of the fuel rods. This will involve calculating the temperature drop through the fuel pellet, gap, clad, and coolant.

It is also important to consider the number of fuel rods required to produce the desired power output, as well as the pressure and coolant flow rate needed to maintain the desired temperature drop. These parameters will help in determining the final design of the reactor core.

In conclusion, the design of a thermal hydraulic system for a nuclear reactor core requires careful consideration of various factors and following a systematic approach to ensure safe and efficient operation. It is important to validate the results using established methods and to continuously monitor and improve the design to optimize its performance.