Nuclear Fuel Pellet Theoretical Density (TD)
Fuel TD adalah nilai densitas fuel pellet pada kondisi ideal/sempurna, yang dihitung dengan persamaan umum berikut:
Fuel TD dihitung dengan asumsi bahwa seluruh volume fuel pellet terisi hanya oleh material fuel, dimana hal ini tidak terlalu tepat, karena sebenarnya di dalam fuel pellet pasti terdapat impuritas, baik itu porositas ataupun rongga2 udara yang sangat kecil, akibat proses fabrikasi yang tidak sempurna. Karena itu densitas fuel yang sebenarnya pasti tidak 100%, melainkan berkisar 94-96% dari fuel TD, dimana fuel TD adalah 10.96 gr/cc.
Nuclear Fuel Pellet Effective Density (ED)
Densitas fuel pellet yang sebenarnya biasa disebut dengan Fuel Pellet Effective Density (ED), dan dirumuskan sebagai berikut:
Nuclear Fuel Pellet Smeared Density (SD)
Fuel SD adalah nilai densitas fuel dengan asumsi bahwa fuel pellet menempati seluruh rongga di dalam cladding, dengan demikian kita asumsikan bahwa fuel pellet menempel dengan permukaan dalam cladding (diasumsikan tidak ada celah/gap). Hubungan antara TD, ED, dan SD adalah sebagai berikut:
Khusus untuk perhitungan cell homogenization dengan menggunakan code PIJ/BURN-SRAC, biasanya digunakan SD, yaitu dengan asumsi bahwa gap tidak terlalu mempengaruhi perhitungan, sehingga dapat diabaikan.
Hydrogen to Heavy Metal Ratio or H/HM for short, is typical parameter in neutronic analysis of a nuclear reactor, or more specifically, the Pressurized Water Reactor (PWR) type. Basically it is simply the ratio of moderator (water) to fuel/fissile material (U/Pu/Th) within one calculation cell. I will briefly show you how to calculate this H/HM.
Consider a standard Westinghouse PWR fuel: cylindrical fuel rod array, arranged in rectangular geometry, please refer to this link for the detail of its technical specifications.
Next, make sure that you have understood how to calculate the atomic number density, or otherwise learn it first from the following link:
Neutron lethargy, or logarithmic energy decrement, u, is a dimensionless logarithm of the ratio of the energy of source neutrons to the energy of neutrons after a collision:
With that definition, the neutron lethargy increases as the neutron slows down, the gain in lethargy after a collision is:
Disclaimer:
Information presented in this article are based on publicly available data of the IRIS reactor project, as properly cited from the original source. This article is NOT part of the official IRIS project led by Westinghouse. For more reliable information, the reader should refer to any official websites and information sources of the IRIS project and/or the IRIS consortium. All trademarks and registered trademarks shown in this article are the property of their respective owners.
General Plant Data
Core thermal power
1000 MWt [ref.2-page35]
Power Plant Net Output
335 MWe [ref.2-page35]
Nuclear Steam Supply System
Number of coolant loops
Integral RCS [ref.2-page35]
Steam temperature/pressure
317/5.8 °C/MPa [ref.2-page35]
Feedwater temperature/pressure
224/6.4 °C/MPa [ref.2-page35]
Reactor Coolant System
Total core flow rate
36000 kg/s [ref.3-page53]
Primary coolant flow rate
4700 kg/s [ref.2-page35]
Reactor operating pressure
15.5 MPa [ref.2-page35]
Core inlet temperature
292 °C [ref.2-page35]
Core (riser) outlet temperature
330 °C [ref.2-page35]
Reactor Core
Fuel assembly total length
5.207 m [ref.2-page35]
Fuel inventory
48.5 tU [ref.2-page35]
Average linear heat rate
10.0 kW/m [ref.2-page35]
Average core power density (volumetric)
51.26 kW/l [ref.2-page35]
Specific power
(= core thermal power/fuel inventory)
20.6186 kW/kg-HM
Fuel material
Sintered UO2 [ref.2-page35]
Westinghouse standard PWR fuel
Fuel average density
96% Theoretical Density [ref.3-page203]
UO2-TD = 10.96 g/cc
Rod array
Square
17×17 XL [ref.2-page38,ref.5-page155]
Number of fuel assemblies
89 [ref.2-page35]
Number of fuel rods/assembly
264 [ref.2-page35]
Fuel pellet diameter
8.19 mm [ref.1-page634]
Pellet-clad gap
0.082 mm [ref.1-page634]
Clad thickness
0.572 mm [ref.1-page634]
Outer diameter of fuel rods
9.5 mm [ref.2-page35,ref.5-page155]
Pitch (center-to-center)
12.54 mm [ref.1-page634]
P/D
1.32 [ref.3-page34]
Average H/HM ratio
(Hydrogen to Heavy Metal ratio)
Here is a little example of 2D Computational Fluid Dynamics (CFD), the basic equation beeing used is the non-compressible Navier-Stokes equation. The code was written in C language, originally written by Professor Takayuki Aoki, TokyoTech GSIC.
download the code here, extract, compile and run it on UNIX/LINUX platform, make sure that gcc and ImageMagick are correctly installed on your system
The core of this freeware is The Fourth Order Runge-Kutta Method (FORK) which numerically solve the coupled differential decay equations. Downloadnucmed.doc , rename it to nucmed.zip, then extract. As usual, this code was written in Pascal under Borland Dephi 7.
Screenshots :
Keywords : runge kutta radioactivity decay nuclear medicine
In this article I want to share a little knowledge about a simple Monte Carlo Method which applied to solve the infinite slab problem. The basic idea of the Monte Carlo Method as it applied to the infinite slab problem is we want to know the distribution of particles that bombarded to the slab, as the particles performing a random walk. We can simulate this phenomena if we know the required physical variables of the system, in this case, the reaction cross sections. For further explanation, first please downloadmontecarlo1.doc , rename it to montecarlo1.zip, then extract. Inside the “montecarlo1″ folder, you can find 2 explanation files, power point slide, and a computer source code. This source code is real implementation of monte carlo model in form of software, this code was written in Pascal Language of Borland Delphi 7.
Solution of One Group Neutron Diffusion Equation in Spherical Geometry With S.O.R Method
Pada artikel ini saya ingin berbagi ilmu mengenai teknik memecahkan persamaan difusi neutron pada geometri bola (sferis), secara numerik dengan menggunakan program komputer. Reaktor nuklir yang akan dianalisis berbentuk bola sempurna dengan radius R, dan kuantitas yang dicari adalah profil distribusi flux neutron. Pertama silahkan download file “difusibola.doc“, kemudian rename menjadi “difusibola.zip”, kemudian ekstrak. Di dalam folder “difusibola” terdapat file “penjelasan.doc” dan folder “program”. File “penjelasan.doc” berisi teori dan penjelasan singkat mengenai persamaan difusi neutron dan solusi numeriknya, sedangkan folder “program” berisi implementasi real-nya berupa source code program komputer yang saya tulis dengan menggunakan bahasa Pascal pada Borland Delphi 7. Bila anda hanya ingin menjalankan programnya saja, dobel klik file “Project1.exe”, bila anda ingin melihat script-nya, dobel klik file “Project1.dpr”.
Solution of One Group-One Dimension Neutron Diffusion Equation
Pada artikel ini saya ingin berbagi ilmu mengenai teknik memecahkan persamaan difusi neutron 1 grup 1 dimensi, secara numerik dengan menggunakan program komputer. Pertama silahkan download file “difusi1g1d.doc“, kemudian rename menjadi “difusi1g1d.zip”, kemudian ekstrak. Di dalam folder “difusi1g1d” terdapat file “penjelasan.doc” dan folder “program”. File “penjelasan.doc” berisi teori dan penjelasan singkat mengenai persamaan difusi neutron dan solusi numeriknya, sedangkan folder “program” berisi implementasi real-nya berupa source code program komputer yang saya tulis dengan menggunakan bahasa Pascal pada Borland Delphi 7. Bila anda hanya ingin menjalankan programnya saja, dobel klik file “Project1.exe”, bila anda ingin melihat script-nya, dobel klik file “Project1.dpr”.
