scholarly journals VALIDATION OF SRAC CODE SYSTEM FOR NEUTRONIC PARAMETERS CALCULATION OF THE PWR MOX/UO2 CORE BENCHMARK

2021 ◽  
Vol 27 (1) ◽  
pp. 47
Author(s):  
Wahid Luthfi ◽  
Surian Pinem
Keyword(s):  
Boron Concentration ◽  
Reactor Core ◽  
Axial Direction ◽  
Code System ◽  
Pressurized Water ◽  
Calculation Results ◽  
Neutronic Parameters ◽  
Weighted Error ◽  
Parameters Calculation

VALIDATION OF SRAC CODE SYSTEM FOR NEUTRONIC PARAMETERS CALCULATION OF THE PWR MOX/UO2 CORE BENCHMARK. Determination of neutronic parameter value is an important part in determining reactor safety, so accurate calculation results can be obtained. This study is focused on the validation of SRAC code system in the calculation of neutronic parameters value of a PWR (Pressurized Water Reactor) reactor core. MOX/UO2 Core Benchmark was choosed because it is used by several researchers as a reference core for code validation in the determination of neutronic parameters of a reactor core. The neutronic parameters calculated include critical boron concentration, delayed neutron fraction, and Power Peaking Factor (PPF) and its distribution in axial and radial directions. When compared with reference data, the calculation results of the critical boron concentration value show that there is a difference of 22.5 ppm on SRAC code system. Meanwhile, differences in power per fuel element (assembly power error) value of power-weighted error (PWE) and error-weighted error (EWE) is 2.93% and 3.94%, respectively. Maximum difference between PPF value in axial direction with reference reaches a value of 4.57%. SRAC calculation results also show consistency with the calculation results of other program packages or code. Results of this study indicate that SRAC code system is still quite accurate for the calculation of neutronic parameters of PWR reactor core benchmark. Therefore, SRAC code system can be used to calculate neutronic parameters of PWR reactor core, especially when using MOX (mixed oxide) fuel.Keywords: Neutronic parameter, critical boron concentration, power peaking factor, SRAC code system.

Pressurized Water Reactor Core Crud Mapping

2010 ◽  
Author(s):  
W. A. Byers
Keyword(s):  
Reactor Core ◽  
Pressurized Water Reactor ◽  
Cleaning Efficiency ◽  
Pressurized Water ◽  
Advantages And Disadvantages ◽  
Radiation Fields ◽  
Video Images ◽  
Core Deposits ◽  
Core Areas

Knowledge of the quantity and distribution of core deposits (crud) can be useful in many ways. Trending the amount of crud can show the effects of coolant chemistry changes and crud remediation efforts. The crud distribution reflects the locations of sub-cooled boiling, and indirectly, core performance. Both the quantity and distribution of crud can be useful in assessing the risk of Crud Induced Power Shifts (CIPS) and Crud Induced Local Corrosion (CILC). Measuring core crud is also essential in increasing the fundamental understanding of crud transport and the build-up of ex-core radiation fields. Several techniques have been used to measure core crud levels. Crud scraping is frequently applied to sample crud, but it cannot be applied economically over broad areas. Crud has been mapped over broad core areas by analyzing the waste stream from an ultrasonic fuel cleaner, but uncertainties in the cleaning efficiency limit the accuracy of this technique. Video crud mapping, the determination of crud coverage from video images of fuel assemblies, is an economical method that can be readily applied to peripheral rods on every fuel assembly. The mapping can be performed by computer image analysis, or by visual comparisons to standards. The two techniques are compared and the advantages and disadvantages of each are discussed. The effectiveness of video crud mapping in assessing fuel performance issues is shown in the results of several field applications.

scholarly journals CALCULATION OF 2-DIMENSIONAL PWR MOX/UO2 CORE BENCHMARK OECD NEA 6048 WITH SRAC CODE

2020 ◽  
Vol 22 (3) ◽  
pp. 89
Author(s):  
Wahid Luthfi ◽  
Surian Pinem
Keyword(s):  
Cross Sections ◽  
Power Distribution ◽  
Reactor Core ◽  
Design Parameters ◽  
Maximum Deviation ◽  
Maximum Difference ◽  
Code System ◽  
Spent Fuel ◽  
Neutronic Parameters ◽  
Safety Characteristic

The mixed uranium-plutonium oxide fuel (MOX/UO2) is an interesting fuel for future power reactors. This is due to the large amount of plutonium that can be processed from spent fuel of nuclear plants or from plutonium weapons. MOX/UO2 fuel is very flexible to be applied in thermal reactors such as PWR and it is more economical than UO2 fuel. However, due to the different nature of neutron interactions of MOX in PWR, it will change the reactor core design parameters and also its safety characteristic. The purpose of this study is to determine the accuracy of SRAC2006 code system in generation of cross-sections and calculation of reactor core design parameters such as criticality, reactivity of control rods and radial power distribution. In this study, PWR MOX/UO2 Core Transient Benchmark is used to verify the code that models a MOX/UO2 fueled core. SRAC-CITATION result is different from DeCART by 0.339% from. SRAC-CITATION result of single rod worth in All Rods Out (ARO) conditions are quite good with a maximum difference of 6.34% compared to BARS code and 4.74% compared to PARCS code. In All Rods In (ARI) condition, SRAC-CITATION results compared to the PARCS code is quite good where the maximum difference is 9.72%, but compared to BARS code, it spikes up to 33.24% at maximum difference. In the other case, overall radial power density results are quite good compared to the reference. Its maximum deviation from DeCART code is 5.325% in ARO condition and 6.234% in ARI condition. Based on the results of these calculations, SRAC code system can be used to generate cross-section and to calculate some neutronic parameters. Hence, it can be used to evaluate the neutronic parameters of the MOX/UO2 PWR core design.Keywords: MOX/UO2 fuel, Criticality, Power peaking factor, SRAC2006

scholarly journals KINETIC PARAMETERS CALCULATION DURING FIRST CYCLE OF THE WWER-1000 REACTOR CORE

2019 ◽  
Vol 8 (1) ◽  
pp. 63-70
Author(s):  
Morteza Akbari ◽  
Samira Rezaei ◽  
Farrokh Khoshahval
Keyword(s):  
Kinetic Parameters ◽  
Probabilistic Method ◽  
Reactor Core ◽  
Neutron Generation ◽  
Delayed Neutron ◽  
Neutron Lifetime ◽  
Reactor Physics ◽  
And Control ◽  
Parameters Calculation

Determination of the effective delayed neutron fraction (βeff) and neutron generation time (Λ), on account of their important role in the reactivity transients analysis, safety, and control of nuclear reactors, is of the great importance in the reactor physics calculations. In this paper, evaluation of the kinetic parameters (effective delayed neutron fraction and prompt neutron lifetime) in PWRs is calculated. Software was developed to automate the procedure of kinetic parameters calculations. We used both a deterministic and a probabilistic method for calculation of the delayed neutron parameters. The results performed well in comparison to the reference.

scholarly journals Determination of the neutron flux in the reactor zones with the strong neutron absorption and leakage

2004 ◽  
Vol 1 (3) ◽  
pp. 99-112
Author(s):  
Vladan Ljubenov ◽  
Miodrag Milosevic
Keyword(s):  
Neutron Flux ◽  
Cross Section ◽  
Reactor Core ◽  
Numerical Procedure ◽  
Experimental Methodology ◽  
Neutron Absorption ◽  
Code System ◽  
Neutron Converter ◽  
Neutron Absorption Cross Section

The procedures for the numerical and experimental determination of the neutron flux in the zones with the strong neutron absorption and leakage are described in this paper. Numerical procedure is based on the application of the SCALE-4.4a code system where the Dancoff factors are determined by the VEGA2DAN code. Two main parts of the experimental methodology are measurement of the activity of irradiated foils and determination of the averaged neutron absorption cross-section in the foils by the SCALE-4.4a calculation procedure. The proposed procedures have been applied for the determination of the neutron flux in the internal neutron converter used with the RB reactor core configuration number 114.

Determination of the Average Energy of Iron Ions in the Axial Direction of a Micropinch Discharge Using an Ion Collector

2019 ◽  
Vol 82 (10) ◽  
pp. 1387-1391
Author(s):  
S. A. Sarantsev ◽  
I. F. Raevskii ◽  
V. A. Kostyushin ◽  
A. S. Savelov
Keyword(s):  
Average Energy ◽  
Axial Direction ◽  
Iron Ions

Optimization of a Specimen Size for the Determination of the Diffusible Hydrogen Content in Metals

2014 ◽  
Vol 698 ◽  
pp. 466-471
Author(s):  
Oleg V. Panchenko ◽  
Alexey M. Levchenko ◽  
Victor A. Karkhin
Keyword(s):  
Hydrogen Content ◽  
Calculation Method ◽  
Specimen Size ◽  
Experimental Results ◽  
Diffusible Hydrogen ◽  
Measuring Methods ◽  
Deposited Metal ◽  
Calculation Results

Specimens of various sizes are used to determine hydrogen content in deposited metals in such standards as ISO 3690, AWS A 4.3, and GOST 23338 while measuring methods are the same. It causes problems in comparison of experimental results and brings up the following question: what kind of specimen size is optimal to determine hydrogen content? An optimal specimen size was estimated using a calculation method. Experimental and calculation results obtained by using specimens with estimated dimensions were compared to the results obtained by using the specimen with dimensions of 100*25*8 mm to determine hydrogen content in a deposited metal.

Neutronic Analysis of Accident-Tolerant Fuel Assemblies With Silicon Carbide Cladding in Pressurized Water Reactors

2017 ◽  
Author(s):  
Zhixiong Tan ◽  
Jiejin Cai
Keyword(s):  
Silicon Carbide ◽  
Fuel Assembly ◽  
Power Distribution ◽  
Safety Margin ◽  
Severe Accident ◽  
Pressurized Water Reactors ◽  
Fuel Pellet ◽  
Pressurized Water ◽  
Neutronic Parameters ◽  
Water Reactors

After Fukushima Daiichi Nuclear Power Plant accident, alternative fuel-design to enhance tolerance for severe accident conditions becomes particularly important. Silicon carbide (SiC) cladding fuel assembly gain more safety margin as novel accident tolerant fuel. This paper focuses on the neutron properties of SiC cladding fuel assembly in pressurized water reactors. Annular fuel pellet was adopted in this paper. Two types of silicon carbide assemblies were evaluated via using lattice calculation code “dragon”. Type one was consisted of 0.057cm SiC cladding and conventional fuel. Type two was consisted of 0.089cm SiC cladding and BeO/UO2 fuel. Compared the results of SiC cladding fuel assembly neutronic parameters with conventional Zircaloy cladding fuel assembly, this paper analyzed the safety of neutronic parameters performance. Results demonstrate that assembly-level reactivity coefficient is kept negative, meanwhile, the numerical value got a relatively decrease. Other parameters are conformed to the design-limiting requirement. SiC kinds cladding show more flat power distribution. SiC cases also show the ability of reducing the enrichment of fuel pellets even though it has higher xenon concentration. These types of assembly have broadly agreement neutron performance with the conventional cladding fuel, which confirmed the acceptability of SiC cladding in the way of neutron physics analysis.

CFD Analysis of the Coolant Mixing within the Upper Plenum of a Pressurized Water Reactor (PWR)

2013 ◽  
Vol 444-445 ◽  
pp. 411-415 ◽  
Author(s):  
Fu Cheng Zhang ◽  
Shen Gen Tan ◽  
Xun Hao Zheng ◽  
Jun Chen
Keyword(s):  
Temperature Distribution ◽  
Fluid Dynamic ◽  
Characteristic Temperature ◽  
Reactor Core ◽  
Flow Distribution ◽  
Sensitivity Study ◽  
Cfd Model ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Water Reactor

In this study, a Computational Fluid Dynamic (CFD) model is established to obtain the 3-D flow characteristic, temperature distribution of the pressurized water reactor (PWR) upper plenum and hot-legs. In the CFD model, the flow domain includes the upper plenum, the 61 control rod guide tubes, the 40 support columns, the three hot-legs. The inlet boundary located at the exit of the reactor core and the outlet boundary is set at the hot-leg pipes several meters away from upper plenum. The temperature and flow distribution at the inlet boundary are given by sub-channel codes. The computational mesh used in the present work is polyhedron element and a mesh sensitivity study is performed. The RANS equations for incompressible flow is solved with a Realizable k-ε turbulence model using the commercial CFD code STAR-CCM+. The analysis results show that the flow field of the upper plenum is very complex and the temperature distribution at inlet boundary have significant impact to the coolant mixing in the upper plenum as well as the hot-legs. The detailed coolant mixing patterns are important references to design the reactor core fuel management and the internal structure in upper plenum.

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