Neutron Physics Characterization & Optimization Analysis of the ACPR100 Small Modular Reactor

2021 ◽  
Author(s):  
Songyang Liu ◽  
Xiang Wang
Keyword(s):  
Life Cycle ◽  
Reactor Core ◽  
Fission Products ◽  
Negative Temperature ◽  
Design Parameters ◽  
Monte Carlo Code ◽  
Time Step ◽  
Optimization Analysis ◽  
Insertion Depth ◽  
Pressurized Water

Abstract The ACPR100 is a small modular pressurized water reactor design proposed by China General Nuclear Power Corporation which integrates most important components of the reactor into one pressure vessel. This paper aims to model the reactor core by Monte Carlo code Serpent. Firstly, the steady-state characterization and optimization analysis within different temperature, pin-pitch and other design parameters are performed. Secondly, the loading pattern of fuel assemblies with Gd-doped fuel rods was assessed. Thirdly, based on the reference research, the temperature coefficients of fuel and coolant are calculated. The influence of control rod insertion depth was simulated additionally. Besides, to show the life cycle and the change of inventory directly, we performed burnup calculation based on pre-defined time step and discussed the fission products including radioactive minor actinides and radiopharmaceutical isotopes during the life cycle. The results show that the ACPR100 achieved more stable performance using high abundance boric acid, and the negative temperature reactivity coefficient is sufficient to maintain the stability of the reactor operation, but the ACPR100 is not suitable for massive production of radiopharmaceutical isotopes.

scholarly journals Actinides induced irradiation damage and swelling effect in irradiated Zircaloy-4 after 30 years of storage

2021 ◽  
Vol 1 ◽  
pp. 7-8
Author(s):  
Mara Marchetti ◽  
Michel Herm ◽  
Tobias König ◽  
Simone Manenti ◽  
Volker Metz
Keyword(s):  
Nuclear Fuel ◽  
Spent Nuclear Fuel ◽  
Reactor Core ◽  
Fission Products ◽  
Alpha Decay ◽  
Irradiation Damage ◽  
Monte Carlo Code ◽  
Pressurized Water ◽  
Dry Storage ◽  
Inner Surface

Abstract. After several years in the reactor core, irradiated nuclear fuel is handled and subsequently stored for a few years under water next to the core, to achieve thermal cooling and decay of very short-lived radionuclides. Thereafter, it might be sent to dry-cask interim storage before final disposal in a deep geological repository. Here, the spent nuclear fuel (SNF) is subject to a series of physicochemical phenomena which are of concern for the integrity of the nuclear fuel cladding. After moving the SNF from wet to dry storage, the temperature increases, then slowly decreases, leading the hydrogen in solid solution in the cladding to precipitate radially with consequent hydride growth and cladding embrittlement (Kim, 2020). Another phenomenon affecting the physical properties of the cladding during interim dry storage is the irradiation damage produced in the inner surface of the cladding by the alpha decay of the actinides present at the periphery of the pellet, particularly when the burnup at discharge is high. SNF pellets with high average burnup present larger fuel volumes at the end of their useful life due to accumulation of insoluble solid fission products and noble gases, which leads to disappearance of the as-fabricated pellet–clad gap. Further swelling is expected as a consequence of actinide decay and the accumulation of helium. This leads to larger cladding hoop stress and larger alpha decay damage. The present work first investigates the variation in diameter caused by pellet swelling in an irradiated Zircaloy-4 cladding after chemical digestion of the uranium oxide (UOx) pellet. Second, the irradiation damage produced during the 30 years elapsed since the end of irradiation in terms of displacements per atom (dpa) is studied by means of the FLUKA Monte Carlo code. The irradiation damage produced by the decay of actinides in the inner surface of the cladding extends for less than 3 % in depth. The considered cladded UOx pellet was extracted from a pressurized water reactor (PWR) fuel rod consisting of five segments, with an average burnup at discharge of 50.4 GWd (tHM)−1.

scholarly journals Burnup simulations of different fuel grades using the MCNPX Monte Carlo code

2014 ◽  
Vol 29 (4) ◽  
pp. 259-267
Author(s):  
Fiifi Asah-Opoku ◽  
Zhihua Liang ◽  
Ziaul Huque ◽  
Raghava Kommalapati
Keyword(s):  
Monte Carlo ◽  
Reactor Core ◽  
Fuel Burnup ◽  
Monte Carlo Code ◽  
Oxide Fuel ◽  
Pressurized Water ◽  
Plutonium Isotopes ◽  
Unequal Distribution ◽  
Mox Fuel ◽  
Enriched Uranium

Global energy problems range from the increasing cost of fuel to the unequal distribution of energy resources and the potential climate change resulting from the burning of fossil fuels. A sustainable nuclear energy would augment the current world energy supply and serve as a reliable future energy source. This research focuses on Monte Carlo simulations of pressurized water reactor systems. Three different fuel grades - mixed oxide fuel (MOX), uranium oxide fuel (UOX), and commercially enriched uranium or uranium metal (CEU) - are used in this simulation and their impact on the effective multiplication factor (Keff) and, hence, criticality and total radioactivity of the reactor core after fuel burnup analyzed. The effect of different clad materials on Keff is also studied. Burnup calculation results indicate a buildup of plutonium isotopes in UOX and CEU, as opposed to a decline in plutonium radioisotopes for MOX fuel burnup time. For MOX fuel, a decrease of 31.9% of the fissile plutonium isotope is observed, while for UOX and CEU, fissile plutonium isotopes increased by 82.3% and 83.8%, respectively. Keff results show zircaloy as a much more effective clad material in comparison to zirconium and stainless steel.

scholarly journals SENSITIVITY ANALYSIS OF STAINLESS STEEL REFLECTOR FOR VR-1 TRAINING REACTOR

2021 ◽  
Vol 247 ◽  
pp. 08003
Author(s):  
Jan Frybort ◽  
Lubomir Sklenka ◽  
Filip Fejt ◽  
Pavel Suk ◽  
Lenka Frybortova
Keyword(s):  
Stainless Steel ◽  
Power Distribution ◽  
Reactor Core ◽  
Sensitivity Study ◽  
Fast Neutrons ◽  
Monte Carlo Code ◽  
Pressurized Water ◽  
The Core ◽  
Water Reactors ◽  
Radial Reflector

Pressurized water reactors are typically surrounded in the radial direction by neutron reflectors made from stainless steel and water. These reflectors decrease neutron leakage and provide protection of pressure vessel from fast neutrons damaging its integrity. Such a radial reflector influences multiplication factor of the core and distribution of neutron flux and fission power inside the core. All these effects can be analyzed by full-core simulations using macroscopic constants. Methodology for generation of the macroscopic constants for non-fuel regions will be tested for new stainless steel reflectors at the VR-1 reactor. Rods from SS 304l material will be used for construction of radial reflectors for the VR-1 reactor. They will be design to generate sufficient measurable response in selected core characteristics. The study is focused on core power distribution and reactivity worth of absorbing rods in a VR-1 reactor core. The core typically consists of about 20 IRT-4M fuel assemblies and seven absorbing rods UR-70. Replacing water surrounding the core by several reflector assemblies containing stainless steel will influence leakage and distribution of neutrons inside the core. The current analysis deals with local effects and employs the sensitivity study to discover the nature of reflectors’ impact on the reactor core. These effects were studied even for several past VR-1 reactor core configurations. All calculations were carried out in Serpent2 Monte-Carlo code with various evaluated libraries: ENDF/B-VII.1, ENDF/B-VIII.0, and JEFF-3.3 data.

The Verification of SONG Library

2017 ◽  
Author(s):  
Hua Bei ◽  
Jinkun Zhao ◽  
Qichang Chen ◽  
Shengyi Si
Keyword(s):  
Layer Structure ◽  
Resonance Effect ◽  
Fission Products ◽  
Reaction Rates ◽  
Monte Carlo Code ◽  
Personal Factors ◽  
Operation Condition ◽  
Pressurized Water ◽  
Fission Rate ◽  
Wide Range

We have developed the 293-group cross-section library SONGLIB with complete reaction pathes, broad-spectrum solving ability and wide range isotopes, which is adopted in the lattice code SONG, the next generation reactor lattice calculation and analysis code. SONGLIB provides the multi-group data for the transport calculation, resonance calculation and burnup calculation of SONG. So the precision and reliability of data is very important. However, the size of the data is very large and there are still a lot of personal factors in determination of the processing parameters and methods. In order to evaluate the library, it’s necessary to carry out the test work comprehensively. Amount of test work is then undertaken based on three kinds of reactors such as pressurized water reactor with UO2 fuel, sodium cooled fast reactor and new-type molten salt reactor with none-beryllium core. In order to test the adaptivity of the library, 39 cases are designed totally, changing the fuel composition, the absorber compositon, the burnup depth and the working temperature. Only the results of a few cases are given in details to shorten the length of this article. For comparison, the Monte Carlo code MCNP is chosen and the library source of evaluated nuclear data is ensured to be the same. The model that has been calculated is simple, the cell with “fuel-cladding-modulator” the three-layer structure and total reflection boundary for the aim of minimizing the differences between the codes. From the macroscopic angle, the results of infinite multiplication factor fit well with each other in total. From the microscopic angle, the absorption rate for actinides, fission products and absorbers, the fission rate for fissile isotopes and scattering rate for moderators are compared with that of MCNP whose output is normalized to the actual cell power. The difference between the reaction rates for most isotopes can be neglected. But there still exist some obvious differences for some isotopes whose resonance effect should have been considered, such as Fe56, ZrNat and W, which might affect the macoscopic results to some extent. As a conclusion, the data of SONGLIB can be expected with high precision and reliability. And SONGLIB is adapted in solving various problems for reactors with different spectrum, depth of burnup, operation condition or fuel cycling. Nevertheless, there still exist some factors that may affect the final results during the process and use of the library, which should be paid attention to while taking a further step in optimizing and updating the library.

Thermodynamic Data of Iodine Reactions Calculated by Quantum Chemistry. Training Set of Molecules

2008 ◽  
Vol 73 (10) ◽  
pp. 1340-1356 ◽  
Author(s):  
Katarína Mečiarová ◽  
Laurent Cantrel ◽  
Ivan Černušák
Keyword(s):  
Quantum Chemistry ◽  
Ab Initio ◽  
Thermodynamic Data ◽  
Theoretical Calculations ◽  
Reactor Core ◽  
Fission Products ◽  
Vapour Phase ◽  
Reactor Accident ◽  
Training Set ◽  
Coolant System

This paper focuses on the reactivity of iodine which is the most critical radioactive contaminant with potential short-term radiological consequences to the environment. The radiological risk assessments of 131I volatile fission products rely on studies of the vapour-phase chemical reactions proceeding in the reactor coolant system (RCS), whose function is transferring the energy from the reactor core to a secondary pressurised water line via the steam generator. Iodine is a fission product of major importance in any reactor accident because numerous volatile iodine species exist under reactor containment conditions. In this work, the comparison of the thermodynamic data obtained from the experimental measurements and theoretical calculations (approaching "chemical accuracy") is presented. Ab initio quantum chemistry methods, combined with a standard statistical-thermodynamical treatment and followed by inclusion of small energetic corrections (approximating full configuration interaction and spin-orbit effects) are used to calculate the spectroscopic and thermodynamic properties of molecules containing atoms H, O and I. The set of molecules and reactions serves as a benchmark for future studies. The results for this training set are compared with reference values coming from an established thermodynamic database. The computed results are promising enough to go on performing ab initio calculations in order to predict thermo-kinetic parameters of other reactions involving iodine-containing species.

Research on the Optimal Design Method of the Main Components for the Wheelchair Pickup

2013 ◽  
Vol 791-793 ◽  
pp. 799-802
Author(s):  
Ya Ping Wang ◽  
H.R. Shi ◽  
L. Gao ◽  
Z. Wang ◽  
X.Y. Jia ◽  
...  
Keyword(s):  
Parametric Design ◽  
Design Method ◽  
Basic Function ◽  
Design Parameters ◽  
Algorithm Optimization ◽  
Optimization Analysis ◽  
Research Designs ◽  
Optimal Design Method ◽  
Main Components

With the increasing of the aging of population all over the world, and With the inconvenience coming from diseases and damage, there will be more and more people using the wheelchair as a tool for transport. When it cant be short of the wheelchair in the daily life, the addition of the function will bring the elevation of the quality of life for the unfortunate. Staring with this purpose, the research designs a pickup with planetary bevel gear for the wheelchair. After determining the basic function of the wheelchair aids, the study determines the design parameters by using the knowledge of parametric design and completes the model for the system with Pro/E, on the other hand, it completes key components optimization analysis which is based on genetic algorithm optimization.

Optimization Analysis of the Wedge Block Surface of NYD Contact Backstop

2013 ◽  
Vol 433-435 ◽  
pp. 2277-2281
Author(s):  
Quan Wei Wang ◽  
Ming Hui Wang ◽  
Dong Li ◽  
Dian Mao Wan ◽  
Rong Meng
Keyword(s):  
Cross Section ◽  
Design Parameters ◽  
Optimization Analysis ◽  
Archimedes Spiral ◽  
The Cross ◽  
Section Curve ◽  
Relationship Of ◽  
The Relationship

By analyzing the relationship of the design parameters of NYD contact backstop, the cross-section curve of the wedge block has been discussed as Archimedes spiral, logarithm spiral and arc. Each curve is designed optimally using MATLAB optimization toolbox. The merits and drawbacks of each curve are discussed.

Sign in / Sign up

Export Citation Format

Share Document