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.

Neutronic and Physical Characteristics of an Accelerator Driven System with a Lead-208 Coolant

2006 ◽  
Author(s):  
Georgy L. Khorasanov ◽  
Anatoly P. Ivonov ◽  
Anatoly I. Blokhin
Keyword(s):  
Proton Beam ◽  
Neutron Source ◽  
Lead Isotope ◽  
External Source ◽  
Beam Current ◽  
Thermal Capacity ◽  
Multiplication Factor ◽  
The Core ◽  
Effective Multiplication Factor ◽  
Effective Multiplication

In the paper a possibility of using a lead isotope, pure Pb-208, as a coolant for a subcritical core of 80 MW thermal capacity of the PDS-XADS type facility is considered. Calculations of neutronic characteristics were performed using Monte Carlo technique. The following initial data were chosen: an annular core with a target, as a neutron source, at its centre; the core coolant — Pb-208 (100%); a fuel — a mix of mono nitrides of depleted uranium and power plutonium with a small share of neptunium and americium; the target coolant — a modified lead and bismuth eutectic, Pb-208(80%)-Bi(20%); proton beam energy — 600 MeV; effective multiplication factor of the core under operation — Keff = 0.97; thermal capacity of the core — N = 80 MW. From calculations performed it follows that in using Pb-208 as the core coolant the necessary intensity of the external source of neutrons to deliver 80 MW thermal capacity is equal to S = 2.29−1017 n/s that corresponds to proton beam current Ip = 2.8 mA and beam capacity Pp = 1.68 MW. In using natural lead instead of Pb-208 as the core coolant, effective multiplication factor of the core in normal operating regime falls down to the value equal to Keff = 0.95. In these conditions multiplication of external neutrons in the core and thermal capacity of the subcritical core are below nominal by 1.55 times. For achievement the rated core power N = 80 MW it is required on ∼20–30% to increase the fuel loading and volume of the core, or by 1.55 times to increase intensity of the external source of neutrons. In the last case, the required parameters of the neutron source and of the corresponding proton beam are following: intensity of the neutron source S = 3.55·1017 n/s., beam current Ip = 4.32 mA, beam capacity Pp = 2.59 MW. To exploit the accelerator with the reduced proton beam current it will be required about 56 tons of Pb-208, as a minimum, for the core coolant. Charges for its obtaining can be recovered at the expense of the economy of the proton accelerator construction cost. In this case, the acceptable price of the lead isotope Pb-208 must be less than $2,860/kg.

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.

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 Analysis Of Doppler Reactivity Coefficient On The Typical Pwr-1000 Reactor With Mox Fuel

KnE Energy ◽  
2016 ◽  
Vol 1 (1) ◽  
Author(s):  
Rokh Madi
Keyword(s):  
Nuclear Reactor ◽  
Reactor Core ◽  
Nuclear Data ◽  
Fuel Temperature ◽  
Safety Level ◽  
The Core ◽  
Mox Fuel ◽  
Calculation Results ◽  
Reactivity Coefficient ◽  
Fuel Elements

<p>Doppler coefficient is defined as a relation between fuel temperature changes and reactivity changes in the nuclear reactor core. Doppler reactivity coefficient needs to be known because of its relation to the safety of reactor operation. This study is intended to determine the safety level of the  typical PWR-1000 core by calculating the Doppler reactivity coefficient in the core with UO<sub>2</sub> and MOX fuels. The  typical PWR-1000 core  is similar to the PWR AP1000 core designed by Westinghouse but without Integrated Fuel Burnable Absorber (IFBA) and Pyrex. Inside the core, there are  UO<sub>2</sub> fuel elements with 3.40 % and 4.45 % enrichment, and MOX fuel elements with 0.2 % enrichment. By its own way, the presence of Plutonium in the MOX fuel will contribute to the change in core reactivity. The calculation was conducted using MCNPX code with the ENDF/B- VII nuclear data. The reactivity change was investigated at various temperatures. The calculation results show that the core reactivity coefficient of both UO<sub>2</sub> and MOX fuel are negative, so that the reactor is operated safely.</p>

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.

scholarly journals Effects of various types of reflectors on the 99Mo production in the VVER-Ts reactor targets

2020 ◽  
Vol 6 (2) ◽  
pp. 89-92
Author(s):  
Oleg Yu. Kochnov ◽  
Pavel A. Danilov
Keyword(s):  
Neutron Flux ◽  
Reactor Core ◽  
Raw Material ◽  
The Core ◽  
Assessment Tasks ◽  
Materials Used ◽  
Core Characteristics ◽  
Water Space ◽  
Reactivity Margin ◽  
Effective Multiplication

The effects from introducing various types of reflectors in the VVR-Ts reactor core on the 99Мо production were analyzed. Earlier the effects of only the beryllium reflector on the VVR-Ts reactor core characteristics, such as reactivity margin, neutron flux in experimental channels, and activity of the accumulated 99Мо, were calculated. The calculations are based on a generated precision model of the core which comprises one experimental channel where targets are irradiated for the 99Мо production. The model was built using the SCALE code. The code allows a fairly broad range of calculations to be performed, from criticality estimation to radiological assessment tasks. As the result of the computational analysis of the model, such characteristics were obtained as the effective multiplication factor, the power density in the 99Мо production targets, the neutron flux in the target raw material, and the quantity of the produced 99Мо after 120 hours of irradiation. The data was compared with the results of similar calculations of the VVR-Ts reactor core parameters. Further, the list of the materials used extensively as the reactor core reflector or moderator was formed based on reference literature. A number of models were obtained and analyzed on its basis, in which the water space on the core periphery was substituted for the investigated materials.

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 EFFECT OF β ON EFFECTIVE MULTIPLICATION FACTOR IN 1/fβ SPECTRUM RANDOM SYSTEM

2021 ◽  
Vol 247 ◽  
pp. 17002
Author(s):  
Shouhei Araki ◽  
Yuichi Yamane ◽  
Taro Ueki ◽  
Totaro Tonoike
Keyword(s):  
Spatial Distribution ◽  
Characteristic Parameter ◽  
Noise Model ◽  
Multiplication Factor ◽  
Weierstrass Function ◽  
Random System ◽  
Continuous Space ◽  
Simple Cubic ◽  
Effective Multiplication Factor ◽  
Effective Multiplication

We investigated the effect of β on effective multiplication factor(keff) in the 1/fβ spectrum random system. The random system was generated by the 1/fβ noise model. The model is a continuous space model based on the Randomized Weierstrass function and describes the component spatial distribution with a power spectrum of 1/fβ, where f and β are the frequency domain variable and the characteristic parameter related to randomness, respectively. In this work, the two-group Monte Carlo calculations were performed to obtain the keff for a simple cubic geometry that consisted of two materials (fuel burned to 12 GWd/t and concrete). A large number of replicas having different spatial distribution and characterized by the representative β values were generated using the model, and the distribution on keff was analyzed. We found the dependency on β of standard deviation, skewness, and kurtosis of keff distribution. This result is expected to help to predict the keff distribution due to the randomizing model.

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