scholarly journals Study of Neptunium, Americium and Protactinium Addition for 300MWth GFR with Uranium Carbide Fuel

2019 ◽  
Vol 2 (2) ◽  
pp. 64
Author(s):  
Ratna Dewi Syarifah ◽  
Alvi Nur Sabrina
Keyword(s):  
Spent Nuclear Fuel ◽  
Reactor Core ◽  
Fissile Material ◽  
Uranium Carbide ◽  
Burnable Poison ◽  
The World ◽  
Homogeneous Core ◽  
Effective Multiplication Factor ◽  
Carbide Fuel ◽  
Effective Multiplication

A study of Neptunium, Americium, and Protactinium addition for GFR 300MWth with Uranium Carbide fuel has been performed. The purpose of this study was to determine the characteristics of addition Neptunium, Americium, and Protactinium in a 300MWth Gas-Cooled Fast Reactor. Neutronics calculation was design by using Standard Reactor Analysis Code (SRAC) version 2006 with data nuclides from JENDL-4.0. Neutronics calculations were initiated by calculating the fuel cell calculation (PIJ calculation) and continued with the reactor core calculation (CITATION calculation). The reactor core calculation used two-reactor core configurations, namely the homogeneous core configuration and heterogeneous core configuration. The Neptunium, Americium, and Protactinium additions were performed after obtaining the optimal condition from heterogeneous core configuration. The addition of Neptunium and Americium which are Spent Nuclear Fuel (SNF) from LWR fuels, aims to reduce the amount of Neptunium and Americium in the world and also to reduce the effective multiplication factor (k-eff) value from the reactor. The results obtained that the addition of Neptunium and Americium causes the k-eff value was decreased at the beginning of burn-up time, but increase at the end of burn-up time. It was because Neptunium and Americium absorb neutrons at the beginning of burn-up time and turns into fissile material at the end of burn-up time. The addition of protactinium in the reactor causes the k-eff value to be decreased both at the beginning of the burn-up time and at the end of the burn-up time. It happens because Protactinium absorbs neutrons both at the beginning of the burn-up time and at the end of the burn-up time. Therefore protactinium is often called a burnable poison.

Analysis of Doppler reactivity of SMART reactor core for hybrid fuel configurations of UO2, MOX and (Th/U)O2 using OpenMC

Kerntechnik ◽  
2021 ◽  
Vol 86 (3) ◽  
pp. 229-235
Author(s):  
Y. Alzahrani ◽  
K. Mehboob ◽  
F. A. Abolaban ◽  
H. Younis
Keyword(s):  
Reactor Core ◽  
Multiplication Factor ◽  
The Core ◽  
Mox Fuel ◽  
The Monte Carlo Method ◽  
Burnable Absorber ◽  
Effective Multiplication Factor ◽  
Reactivity Coefficient ◽  
Increasing Temperature ◽  
Effective Multiplication

Abstract In this study, the Doppler reactivity coefficient has been investigated for UO2, MOX, and (Th/U)O2 fuel types. The calculation has been carried out using the Monte Carlo method ( OpenMC). The effective multiplication factor keff has been evaluated for three materials with four different configurations without Integral Fuel Burnable Absorber (IFBA) rods and soluble boron. The results of MOX fuel, homogenous and heterogeneous thorium fuel configuration has been compared with the core of the reference fuel assembly (UO2). The calculation showed that the effective multiplication factor at 1 000 K was 1.26052, 1.14254, 1.22018 and 1.23771 for reference core, MOX, homogenous and heterogeneous configurations respectively. The results shows that reactivity has decreased with increasing temperature while the doppler reactivity coefficient remained negative. Moreover, the use of (Th/U)O2 homogenous and heterogeneous configuration had shown an improved response compared to the reference core at 600 K and 1 000 K. The doppler reactivity coefficient has been found as –8.98E-3 pcm/K, -0.8 655 pcmK for the homogenous and –8.854 pcm/K, -1.2253 pcm/K for the heterogeneous configuration. However, the pattern remained the same as for the reference core at other temperature points. MOX fuel has shown less response compared to the other fuel configuration because of the high resonance absorption coefficient of Plutonium. This study showed that the SMART reactor could be operated safely with investigated fuel and models.

scholarly journals Desain High Temperature Gas-Cooled Reactor (HTGR) Model Geometri Heksagonal Dua Dimensi dengan Bahan Bakar Thorium Hasil Daur Ulang

2020 ◽  
Vol 1 (1) ◽  
pp. 12-16
Author(s):  
Mutia Utari ◽  
◽  
Yanti Yulianti ◽  
Agus Riyanto ◽  
◽  
...  
Keyword(s):  
High Temperature ◽  
Power Density ◽  
Reactor Core ◽  
Two Dimensions ◽  
Maximum Power Density ◽  
Core Size ◽  
Helium Gas ◽  
Effective Multiplication Factor ◽  
Effective Multiplication ◽  
The Ideal

The Research about the design of high temperature helium gas-cooled reactor (HTGR) terraces with thorium fuel recycled using the SRAC program has been completed. This research includes the percentage of fuel enrichment, reactor core size, reactor core configuration, criticality, and the distribution of the power density. The calculation of reactor core is done in two dimensions \sfrac{1}{6} hexagonal terrace section with a triangular mesh. The fuel is used, i.e. thorium with a burn-up of 20 GWd/t and 30 GWd/t, and helium gas as a cooler. The results obtained in this study show that the ideal HTGR reactor core design with reactor core size and configuration are (x) 22 cm at point (y) = 2035,05 cm and at (y) 11 cm at point (x) = 2035,05 cm, then enrichment in fuel 8%. The result of maximum power density is 550.3685 Watt/cm3 where the position at (x) = 22 cm and axis (y) = 11 with the effective multiplication factor value keff of 1,0000002.

scholarly journals VALIDATION OF PWR-FUEL CODE FOR STATIC PARAMETERS IN THE LWR CORE BENCHMARK

2018 ◽  
Vol 20 (3) ◽  
pp. 111 ◽  
Author(s):  
Iman Kuntoro ◽  
Surian Pinem ◽  
Tagor Malem Sembiring
Keyword(s):  
Reference Values ◽  
Finite Difference Methods ◽  
Reactor Core ◽  
Maximum Deviation ◽  
Multiplication Factor ◽  
Static Parameter ◽  
Calculation Results ◽  
Effective Multiplication Factor ◽  
Different Characteristics ◽  
Effective Multiplication

The PWR-FUEL code is a multi dimensional, multi group diffusion code with nodal and finite difference methods. The code will be used to calculate the fuel management of PWR reactor core. The result depends on the accuracy of the codes in producing the core effective multiplication factor and power density distribution. The objective of this research is to validate the PWR-FUEL code for those cases. The validation are carried out by benchmarking cores of IAEA-2D, KOERBERG-2D and BIBLIS-2D. The all three cases have different characteristics, thus it will result in a good accuracy benchmarking. The calculation results of effective multiplication factor have a maximum difference of 0.014 %, which is greater than the reference values. For the power peaking factor, the maximum deviation is 1.75 % as compared to the reference values. Those results show that the accuracy of PWR-FUEL in calculating the static parameter of PWR reactor benchmarks are very satisfactory.Keywords: Validation, PWR-FUEL code, static parameter. VALIDASI PROGRAM PWR-FUEL UNTUK PARAMETER STATIK PADA TERAS BENCHMARK LWR. Program PWR-FUEL adalah program difusi multi-dimensi, multi-kelompok dengan metode nodal dan metode beda hingga. Program ini akan digunakan untuk menghitung manajemen bahan bakar teras reaktor PWR. Akurasi manajemen bahan bakar teras PWR tergantung pada akurasi program dalam memprediksi faktor multiplikasi efektif teras dan distribusi rapat daya. Untuk itu dilakukan validasi program PWR-FUEL sebagai tujuan dalam penelitian ini.  Validasi PWR-FUEL dilakukan menggunakan teras benchmark IAEA-2D, KOERBERG-2D dan BIBLIS-2D. Ketiga kasus ini mempunyai karaktristik yang berbeda sehingga akan memberikan hasil benchmark yang akurat. Hasil perhitungan faktor multiplikasi efektif terdapat perbedaan maksimum adalah 0,014 % lebih besar dari referensi. Sedangkan untuk perhitungan faktor puncak daya, terdapat perbedaan maksimum 1,75 % dibanding harga referensi. Hasil perhitungan menunjukkan bahwa akurasi paket program PWR-FUEL dalam menghitung parameter statik benchmark reaktor PWR menunjukkan hasil yang sangat memuaskan.Kata kunci: Validasi, program PWR-FUEL, parameter statik

Representative neutronic characteristics calculations for the VVER-1000 reactors using SRAC and MCNP5

2016 ◽  
Vol 6 (2) ◽  
pp. 21-30
Author(s):  
Huu Tiep Nguyen ◽  
Viet Phu Tran ◽  
Tuan Khai Nguyen ◽  
Vinh Thanh Tran ◽  
Minh Tuan Nguyen
Keyword(s):  
Monte Carlo ◽  
Power Distribution ◽  
Reactor Core ◽  
Multiplication Factor ◽  
Benchmark Problem ◽  
Fuel Assemblies ◽  
Effective Multiplication Factor ◽  
Neutronic Characteristics ◽  
Effective Multiplication ◽  
Good Agreement

This paper presents the results of neutronic calculations using the deterministic and Monte-Carlo methods (the SRAC and MCNP5codes) for the VVER MOX Core Computational Benchmark Specification and the VVER-1000/V392 reactor core. The power distribution and keff value have been calculated for a benchmark problem of VVER core. The results show a good agreement between the SRAC and MCNP5 calculations. Then, neutronic characteristics of VVER-1000/V392 such as power distribution, infinite multiplication factor (k-inf) of the fuel assemblies, effective multiplication factor keff, peaking factor and Doppler coefficient were calculated using the two codes.

A Monte Carlo study on burnup treatment in sodium-cooled reactor with Th fuel

Kerntechnik ◽  
2021 ◽  
Vol 86 (4) ◽  
pp. 302-311
Author(s):  
M. E. Korkmaz ◽  
N. K. Arslan
Keyword(s):  
Monte Carlo ◽  
Cross Sections ◽  
Power Distribution ◽  
Reactor Core ◽  
Monte Carlo Study ◽  
Minor Actinides ◽  
Monte Carlo Code ◽  
Effective Multiplication Factor ◽  
Conceptual Core ◽  
Effective Multiplication

Abstract Sodium Cooled Reactors is one of the Generation-IV plants selected to manage the long-lived minor actinides and to transmute the long-life radioactive elements. This study presents the comparison between two-designed SFR cores with 600 and 800 MWth total heating power. We have analyzed a conceptual core design and nuclear characteristic of SFR. Monte Carlo depletion calculations have been performed to investigate essential characteristics of the SFR core. The core calculations were performed by using the Serpent Monte Carlo code for determining the burnup behavior of the SFR, the power distribution and the effective multiplication factor. The neutronic and burn-up calculations were done by means of Serpent-2 Code with the ENDF-7 cross-sections library. Sodium Cooled Fast Reactor core was taken as the reference core for Th-232 burnup calculations. The results showed that SFR is an important option to deplete the minor actinides as well as for transmutation from Th-232 to U-233.

scholarly journals Analisis Burn Up pada Reaktor Cepat Berpendingin Gas Menggunakan Bahan Bakar Uranium Alam

2015 ◽  
Vol 7 (2) ◽  
pp. 78-86
Author(s):  
Feriska Handayani Irka ◽  
Zaki Su’ud
Keyword(s):  
Volume Fraction ◽  
Power Level ◽  
Reactor Core ◽  
Natural Uranium ◽  
Excellent Performance ◽  
Uranium Fuel ◽  
Result Show ◽  
Diameter Core ◽  
Effective Multiplication Factor ◽  
Effective Multiplication

Burn up analysis of gas cooled fast reactor (GCFR) with natural uranium fuel has been done. Burn up modification used in order to make reactor can be operated with natural uranium without enrichment. The reactor core subdivided into 10 regions with the same volume in radial directions. Optimization evaluated by burning natural uranium for 100 years and put each of its burn up result per year in reactor with certain configuration. After 10 years burn up period, fuel from first region was shuffling radially to second region and so on fuel from 9th  shuffling to 10th and then fuel from 10th was carried out from reactor core and fresh uranium input to the first region. Calculation has been done by using SRAC system code with JENDL-32 as library, with cylindrical two dimensional R-Z core models. Shuffling method was used in order to make reactor can be operated using natural uranium. . This natural uranium initially being burned by guessed power level of burn up. The height and  diameter core are 350 cm and 240 cm respectively. The volume fraction for this design is 65% fuel, 10% cladding and 25% coolant; with output power 700 MWTh.  The result show that reactor demonstrated excellent performance with effective multiplication factor 1,055

Determination of βeff for ININ's TRIGA Reactor

2020 ◽  
Vol 6 (3) ◽  
Author(s):  
J. Galicia-Aragón ◽  
R. Raya-Arredondo ◽  
H. S. Cruz-Galindo
Keyword(s):  
Fuel Element ◽  
Reactor Core ◽  
Operation Time ◽  
Multiplication Factor ◽  
Time Period ◽  
Positive Reactivity ◽  
Training Research ◽  
Effective Multiplication Factor ◽  
Effective Multiplication

Abstract The value of βeff for Training Research Isotopes of General Atomics (TRIGA) Mark III reactor, belonging to the National Institute of Nuclear Researches (ININ), is reported. The TRIGA Mark III reactor core was simulated with MCNP6 to deduce the effective multiplication factor (keff) for critical state and after a small insertion of positive reactivity (∼0.20 $). To perform more realistic simulations, we had to incorporate in the composition of the low-enrichment uranium (LEU) fuel element the produced poisons in a time period of six years, considering the operation time in days, during which the reactor was operating at maximum power (1 MWth). The calculation of the βeff value was obtained with the keff results, calculated with the code, and the reactor periods measured experimentally. We also obtained directly the βeff value, through the card of MCNP6 to calculate keff (KOPTS) card of the MCNP6 code in order to compare both values.

scholarly journals The Effects of the Number of Coated Fuel Particles on the Neutronic Aspects of 25 MWt Pebble Bed Reactor with Thorium Fuel

2021 ◽  
Vol 7 (1) ◽  
pp. 102
Author(s):  
Dwi Irwanto ◽  
Nining Yuningsih
Keyword(s):  
Fuel Particle ◽  
Fissile Material ◽  
Particle Fraction ◽  
Pebble Bed ◽  
Pebble Bed Reactor ◽  
Effective Multiplication Factor ◽  
Fuel Particles ◽  
Coated Fuel Particle ◽  
Neutronic Characteristics ◽  
Effective Multiplication

High-Temperature Gas Reactor (HTGR) is a type of reactor that continues to be developed because of its advantages in terms of economic aspects, proliferation resistance, and safety aspects. One of the safety aspect improvements is due to the use of the Coated Fuel Particle (CFP). A coated fuel particle is a fuel with a diameter smaller than 1 mm and is protected by several carbon layers. In the Pebble Bed Reactor (PBR) type of HTGR design, the CFP is placed in a 6 cm fuel ball. How much CFP is put into the fuel ball will determine the neutronic characteristics of the reactor. In this study, the effect of the amount of CFP in the fuel ball on the 25 MWt PBR design using Thorium fuel and its impact on several important neutronic aspects, such as the effective multiplication factor, the amount of fuel enrichment, the utilization of fissile material, and the density of the fissile material formed. The calculation was performed by the Monte Carlo MVP / MVP-BURN code. This study found that the coated fuel particle fraction of 15% was the optimum value for the studied neutronic parameters.

scholarly journals Effect of MOX fuel and the ENDF/B-VIII on the AP1000 neutronic parameters calculations by using MCNP6

2019 ◽  
Vol 34 (4) ◽  
pp. 325-335
Author(s):  
Sonia Reda ◽  
Ibrahim Gomaa ◽  
Ibrahim Bashter ◽  
Esmat Amin
Keyword(s):  
Reactor Core ◽  
Delayed Neutron ◽  
Optimization Criteria ◽  
Mox Fuel ◽  
235U Enrichment ◽  
Reactivity Coefficients ◽  
Effective Multiplication Factor ◽  
Neutronic Parameters ◽  
Ap1000 Reactor ◽  
Effective Multiplication

The present work studies the effect of introducing MOX fuel on Westinghouse AP1000 neutronic parameters. The neutronic calculations were performed by using the MCNP6 code with the ENDF/B-VII.1 library and the new release of the ENDF/B-VIII, the AP1000 core with three 235U enrichment zones (2.35 %, 3.40 %, and 4.45 %). The obtained results showed that the simulated model for the AP1000 core satisfies the optimization criteria as a Westing- house reference. The results which included: effective multiplication factor, keff, delayed neutron fraction, beff, excess reactivity, rex, shutdown margin, temperature reactivity coefficients, whole core depletion, neutron flux, power peaking factor and core power density, were calculated and compared with the available published results. The keff in the cold zero power was found to be 1.20495 and 1.20247 with the ENDF/B-VII.1 and the ENDF/B-VIII libraries, respectively, which matches the value of 1.205 presented in the AP1000 Design Control Document for the UO2 fuel core. On the other hand, keff in the cold zero power was found to be 1.19988 and 1.19860 for MOX core with the ENDF/B-VII.1 and the ENDF/B-VIII libraries, respectively, which show good reception and confirm the safety of the design and efficient modeling of AP1000 reactor core.

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