The implication of Thorium fraction on neutronic parameters of pebble bed reactor

Thorium abundance in the Earth's crust is estimated to be three to four times higher than uranium. This is one potential advantage of Thorium as a provider of attractive fuel to produce nuclear energy. Fewer transuranics produced by Thorium during the fuel burn up in the reactor may also be another advantage for reducing the long-term burden of high-level long-lived waste. The scope of this paper is to study the implication of Thorium fraction on neutronic parameters of pebble bed reactor. The reactor model of HTR-10 was selected, and the (Th, 235U)O2 fuel was used in this study. The MCNP6 code was applied to solve a series of neutron transport calculations with various Thorium fractions in (Th,235U)O2 fuel based on the ENDF/BVII library. The calculation results show that the total temperature coefficient of reactivity of Thorium-added pebble bed reactors is generally more negative than those of LEU-fuelled one, except for 10% Thorium fraction. The kinetic parameters, especially prompt neutron lifetime and neutron generation time of pebble bed reactors, are higher, which means the addition of Thorium in the fuel makes the reactor more easily controlled. However, the burn-up calculations show that the introduction of Thorium in the same fuel kernel as LEU within the pebble bed reactor is unable to lengthen the fuel residence time, except for a minimum of 40% Thorium fraction.

Temperature adjusted cross section libraries used for criticality calculations

The change in the temperature of the nuclear reactor components (fuel, moderator, coolant, and structural materials) is considered to be a significant source of reactivity variation. This change must be taken in account during criticality calculations for safety analysis of the reactor. Hence, the exact representation of temperature in the calculations is very important. In this paper, two PWR assemblies are simulated, solid 16 ⨯ 16 and annular 12 ⨯ 12 fuel assemblies. The infinite multiplication factor and its temperature dependent parameters are calculated for both fuel assemblies. Adjusted temperature dependent libraries are created using makxsf code to exactly represent the different temperature values used in the calculations. It is shown that the results obtained using adjusted cross section libraries are more reliable. The two fuel assembly types follow the same behavior despite the differences in their geometrical configuration. The introduction of annular fuel has a very small effect on the investigated neutronic parameters because the moderator to fuel ratio is preserved.

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

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.

DESIGN OF NEUTRONIC PARAMETERS OF MTR REACTOR USING WIMSD-5B/BATAN-FUEL CODES

BATAN has three aging research reactors, so it is necessary to design a new, more modern MTR type reactor using high-density, low enrichment uranium molybdenum fuel. The thermal neutron flux at the irradiation position is an important concern in the design of research reactors. This analysis is performed using standard computer codes WIMSD-5B and Batan-FUEL. The purpose of this study is to analyze the effect of the core configuration with safety control rods and neutronic parameters using the diffusion method calculation. The reactor core consists of 16 fuel elements and four control rods placed in the 5 x 5 position of the grid plate and is loaded the reflector elements outside the core. The cycle length is also a concern, not less than 20 days, and the reactor can be operated safely with a power of 50 MW. The calculation results show that for the highest fuel loading, which is 450 grams of U7Mo/Al fuel with D2O as a reflector, it will provide the lowest thermal neutron flux at the center of the core irradiation position, namely 1.0 x1015 n/cm2s. The core fuel cycle length will be up to 39 days, meeting the expected acceptance and safety criteria.

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

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

Homogenization Effects of VVER-1000 Fuel Assembly on Criticality Calculations

Due to cost effective and simplicity homogeneous reactors have been widely used for experimental and research purposes. Parameters which are difficult to get from a heterogeneous reactor system can be easily obtained from a homogeneous reactor system and can be applied in the heterogeneous reactor system if the major parametric differences are known. In this study, homogenization effects of VVER (Water Water Energetic Reactor)-1000 fuel assembly on neutronic parameters have been analyzed with the universal probabilistic code MCNP (Monte Carlo N-Particle). The infinite multiplication factor (k∞) has been calculated for the reconfigured heterogeneous and homogenous fuel assembly models with 2 w/o U-235 enriched fuel at room temperature. Effect of mixing soluble boron into the moderator/coolant (H2O) has been investigated for both models. Direct and fission detected thermal to higher energy neutron ratio also has been investigated. Relative power distributions of both models have been calculated at critical and supercritical states. Burnup calculations for both the reconfigured cores have been carried out up to 5 years of operation. Effective delayed neutron fraction (βeff) and prompt removal lifetime (ℓ) also have been evaluated. All the results show significant differences between the two systems except the average relative power.