scholarly journals Neutronic analysis for core conversion (HEU–LEU) of the low power research reactor using the MCNP4C code

Nukleonika ◽  
2015 ◽  
Vol 60 (2) ◽  
pp. 367-371 ◽  
Author(s):  
Saadou Aldawahra ◽  
Kassem Khattab ◽  
Gorge Saba
Keyword(s):  
Research Reactor ◽  
Outer Diameter ◽  
Multiplication Factor ◽  
Delayed Neutron ◽  
Fuel Conversion ◽  
Control Rod ◽  
Safety Parameters ◽  
Effective Multiplication Factor ◽  
Mcnp4c Code ◽  
Effective Multiplication

Abstract Comparative studies for conversion of the fuel from HEU to LEU in the miniature neutron source reactor (MNSR) have been performed using the MCNP4C code. The HEU fuel (UAl4-Al, 90% enriched with Al clad) and LEU (UO2 12.6% enriched with zircaloy-4 alloy clad) cores have been analyzed in this study. The existing HEU core of MNSR was analyzed to validate the neutronic model of reactor, while the LEU core was studied to prove the possibility of fuel conversion of the existing HEU core. The proposed LEU core contained the same number of fuel pins as the HEU core. All other structure materials and dimensions of HEU and LEU cores were the same except the increase in the radius of control rod material from 0.195 to 0.205 cm and keeping the outer diameter of the control rod unchanged in the LEU core. The effective multiplication factor (keff), excess reactivity (ρex), control rod worth (CRW), shutdown margin (SDM), safety reactivity factor (SRF), delayed neutron fraction (βeff) and the neutron fluxes in the irradiation tubes for the existing and the potential LEU fuel were investigated. The results showed that the safety parameters and the neutron fluxes in the irradiation tubes of the LEU fuels were in good agreements with the HEU results. Therefore, the LEU fuel was validated to be a suitable choice for fuel conversion of the MNSR in the future.

scholarly journals Neutronic analysis of control rod effect on safety parameters in Tehran Research Reactor

2018 ◽  
Vol 50 (7) ◽  
pp. 1017-1023 ◽  
Author(s):  
Mina Torabi ◽  
A. Lashkari ◽  
Seyed Farhad Masoudi ◽  
Somayeh Bagheri
Keyword(s):  
Research Reactor ◽  
Control Rod ◽  
Tehran Research Reactor ◽  
Safety Parameters

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 OPTIMIZATION OF FUEL RELOADING PATTERNS FOR A RESEARCH REACTOR BY SIMULATED ANNEALING

2011 ◽  
Vol 14 (1) ◽  
pp. 63-71
Author(s):  
Binh Quang Do
Keyword(s):  
Simulated Annealing ◽  
Optimization Problem ◽  
Research Reactor ◽  
Diffusion Theory ◽  
Computer Code ◽  
Decision Variable ◽  
Safety Parameters ◽  
Fortran Language ◽  
Dalat Research Reactor ◽  
Effective Multiplication

This article presents results obtained from a research into an application of simulated annealing method to the in-core fuel reloading pattern optimization for a research reactor. The decision variable of the optimization problem is a fuel reloading pattern for the next cycle after the present cycle finishes. The objective function maximizes the effective multiplication factor keff at the beginning of cycle while it is established to include an important safety paramater – the power peaking factor, in search process. A procedure for searching the optimal solutions was formed and a computer code was developed in the Fortran language running on PCs. Nuclear safety parameters for the optimization problem are provided from the results of the multigroup neutron diffusion theory computation program CITATION. A sample calculation was performed to find the optimal fuel reloading patterns for the second cycle of the Dalat research reactor and the results are presented in this article.

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 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

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.

scholarly journals Performance optimization of radioactive waste transmutation in accelerator driven system

2018 ◽  
pp. 210-215 ◽  
Author(s):  
Anna Golovkina ◽  
Dmitri Ovsyannikov ◽  
Sorin Olaru
Keyword(s):  
Performance Optimization ◽  
Volume Fraction ◽  
Initial Conditions ◽  
Optimization Approach ◽  
Multiplication Factor ◽  
Atomic Concentration ◽  
Driven System ◽  
Accelerator Driven System ◽  
Effective Multiplication Factor ◽  
Effective Multiplication

Long-lived actinides transmutation in accelerator driven system (ADS) is considered in this paper. The objective is to improve the transmutation performance from the overall radiotoxity of nuclear waste point of view by means of an optimization approach based on control theory. Parameters subjected to determination are the initial values of actinides nuclear concentration in the loaded fuel. The change in time of actinides’ atomic concentration is described by a system of ordinary differential equations with initial conditions, that can be given with some inaccuracy. For such dynamical system a gradient-based optimization algorithm taking into account robustness of initial loading will be constructed. Moreover, constraint on value of effective multiplication factor is considered during numerical optimization. Based on this approach initial atomic concentration values providing the best indicators of actinides burnup can be obtained. Calculation results of actinides volume fraction and effective multiplication factor dynamics are presented.

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.

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