Resonance Calculation Code UFOP Based on the Hyper-Fine Group Neutron Resonance Calculation Method

2010 ◽  
Author(s):  
Yulong Qin ◽  
Hongchun Wu ◽  
Liangzhi Cao ◽  
Qingjie Liu
Keyword(s):  
Calculation Method ◽  
Resonance Energy ◽  
Collision Probability ◽  
Neutron Resonance ◽  
Probability Method ◽  
Calculation Error ◽  
Reactor Physics ◽  
Slowing Down ◽  
Energy Group ◽  
Calculation Code

Resonance self-shielding calculation is very important in reactor physics calculation. Conventional resonance calculation method has some fundamental defects, which hinders its application in some problems. The Hyperfine Energy Group Resonance Calculation Method is studied in this paper and a code named UFOP is developed based on this method. In this method, the resonance energy range is divided into hyperfine energy intervals (tens of thousands) and the collision probabilities are calculated. Then the slowing-down equation is directly solved based on CPM (collision probability method). Some techniques are applied in solving the slowing-down equation for improving computational efficiency and reducing calculation error. A resonance benchmark problem with homogeneous and infinite material is calculated to validate the accuracy of the computation code and the hyper-fine group cross-section library utilized in the code. A PWR fuel cell is also calculated and the results are compared with MCNP. The results show good accuracy of this method and the validity of UFOP code.

Calculation of the Macroscopic Group Constant of Pebble Bed Fluoride Salt Cooled High Temperature Reactor Based on the Collision Probability Method

2017 ◽  
Author(s):  
Dida Zhang ◽  
Guobin Jia ◽  
Long He ◽  
Jiajie Shen ◽  
Zhichao Zhan ◽  
...  
Keyword(s):  
High Temperature ◽  
Complex Structure ◽  
Collision Probability ◽  
Probability Method ◽  
Fluoride Salt ◽  
Pebble Bed ◽  
Slowing Down ◽  
Group Constant ◽  
High Temperature Reactor ◽  
Integral Transport

The pebble bed fluoride salt cooled high temperature reactor (PB-FHR) is one of the generation IV nuclear reactors, a lot of study has concentrated on PB-FHR neutronics all over the world. As the most important part in the study work, the macroscopic group constant must be well prepared. The fuel pebble was chosen for the candidate of PB-FHR due to its outstanding, but the double heterogeneous due to its complex structure causes much difficult in the calculation of macroscopic group constant. In this work, an analytical program named Z2D is written to calculate the macroscopic group cross section. In the program, the collision probability method (CPM) was applied to solve the slowing-down integral transport equation, and the macroscopic constant was evaluated with the obtained neutron flux. Also, the recurrence method was introduced to accelerate the computing speed of slowing-down source. The results were compared with those calculated by MCNP, and good agreements were obtained.

scholarly journals Refinements of Pin-Based Pointwise Energy Slowing-Down Method for Resonance Self-Shielding Calculation-II: Verifications

2021 ◽  
Vol 9 ◽  
Author(s):  
Wonkyeong Kim ◽  
Sooyoung Choi ◽  
Deokjung Lee
Keyword(s):  
Temperature Distribution ◽  
Cross Sections ◽  
Reaction Rates ◽  
Collision Probability ◽  
Fuel Pellet ◽  
Test Problems ◽  
Probability Method ◽  
Fuel Temperature ◽  
Slowing Down ◽  
The Difference

The pin-based pointwise energy slowing-down method (PSM) has been refined through eliminating the approximation for using the pre-tabulated collision probability during the slowing-down calculation. A collision probability table is generated by assuming that material composition and temperature are constant in the fuel pellet using the collision probability method (CPM). Refined PSM (PSM-CPM), which calculates the collision probability in the isolated fuel pellet during the slowing-down calculation using CPM, can consider nonuniform material and temperature distribution. For the methods, the extensive comparative analysis is performed with problems representing various possible conditions in a light water reactor (LWR) design. Conditions are categorized with the geometry, material distribution, temperature profile in the fuel pellet, and burnup. With test problems, PSMs (PSM and PSM-CPM) have been compared with conventional methods based on the equivalence theory. With overall calculation results, PSMs show the accuracy in the eigenvalue with differences in the order of 100 pcm compared to the reference results. There was no noticeable difference in the multigroup cross sections, reaction rates, and pin power distributions. However, PSM-CPM maintains the accuracy in the calculation of the fuel temperature coefficient under the condition with 200% power and nonuniform temperature distribution in the fuel pellet. PSM shows the difference in the eigenvalue in the order of 2,000 pcm for the fictitious pin-cell problem with highly steep temperature profiles and material compositions, but PSM-CPM shows the difference in the eigenvalue within 100 pcm.

Validation of a New Lattice Physics Code GALAXY for PWRs

2010 ◽  
Author(s):  
Kazuya Yamaji ◽  
Hiroki Koike ◽  
Daisuke Sato ◽  
Shinobu Tsubota ◽  
Hideki Matsumoto
Keyword(s):  
Monte Carlo ◽  
Calculation Method ◽  
Post Irradiation ◽  
Energy Group ◽  
Continuous Energy ◽  
The Galaxy ◽  
Critical Experiment ◽  
The Many ◽  
Nuclear Constants ◽  
Calculation Code

A new lattice physics and assembly calculation code GALAXY with the 172 energy-group ENDF/B-VII.0 library has been developed. GALAXY generates few group nuclear constants used in a new core simulator COSMO-S. The GALAXY code uses the many enhanced calculation method for more explicit treatment of neutronics characteristics. The outline of enhanced methods used in GALAXY and the qualification results are shown in this paper. From the qualifications in the continuous energy Monte Carlo benchmark, critical experiment analyses and post irradiation examination (PIE) analyses, GALAXY with the library was validated and the applicability of GALAXY to PWR nuclear design was confirmed.

Study of neutron resonance parameters using slowing down time lead spectrometer

1968 ◽  
Vol 60 (1) ◽  
pp. 113-115
Author(s):  
M.P. Navalkar ◽  
K. Chandramoleshwar ◽  
D.V.S. Ramkrishna
Keyword(s):  
Neutron Resonance ◽  
Resonance Parameters ◽  
Slowing Down ◽  
Time Lead

scholarly journals Calculation of Collision Probability Matrix of Nuclear Fuel Cell as a Function of Neutron Energy Group Using Flat Flux Model

2021 ◽  
Vol 15 ◽  
pp. 121-126
Author(s):  
Mohammad Ali Shafii ◽  
Dian Fitriyani ◽  
Seni H J Tongkukut ◽  
Zaki Su’ud
Keyword(s):  
Boundary Condition ◽  
Fuel Cell ◽  
Nuclear Fuel ◽  
Neutron Transport ◽  
Collision Probability ◽  
Cell Region ◽  
Energy Group ◽  
The Matrix ◽  
Flux Model ◽  
Matrix Calculation

One of the methods that widely used in solving neutron transport equations in the nuclear fuel cell is the collision probability (CP) method. The neutron transport is very important to solve because the neutron distribution is related to the reactor power distribution. The important thing in the CP method is the CP matrix calculation, better known as has an important role in determining the neutron flux distribution in the reactor core. This study uses a linear flat flux model in each cell region for each energy group with white boundary condition. Although the type of reactor used in this study is a fast reactor, the matrix calculation still carried out in fast and thermal group energy. The matrix depends on the number of mesh in each cell region. The matrix formed from the mesh distribution will produce a matrix for each energy group. Because the boundary condition of the system is assumed that there are no contributions neutron source from the outside, the sum of the matrix must be less than one. In general, the results of the calculations in this study are following the theory

The Collision Probability Method in Today’s Computer Environment

2016 ◽  
Vol 182 (4) ◽  
pp. 395-416 ◽  
Author(s):  
Dimitar Altiparmakov ◽  
Robert Wiersma
Keyword(s):  
Collision Probability ◽  
Probability Method ◽  
Computer Environment
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