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

Reda, Sonia M.; Gomaa, Mohammed; Bashter, I.I.; Amin, A Esmat

doi:10.2298/ntrp190705002r

Cited by 7 publications

(5 citation statements)

References 6 publications

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“…NODAL3 code requires cross-section data for fuels and other materials in the reactor core. The crosssection data play a very important role in obtaining accurate neutronic calculation results [5,6]. To un-derstand the effect of using different cross-section data on neutronic parameters, the PWR MOX/UO 2 Core Transient Benchmark case was chosen to be evaluated [7].…”

Section: Introductionmentioning

confidence: 99%

Effect of cross-sections data on calculated static neutronic parameters of PWR MOX/UO2 core transient benchmark case using NODAL3 code

Luthfi,

Pinem

2023

Acta Polytech

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This paper describes the effect of cross-section data generated by several codes on calculated neutronic parameters. The Pressurized Water Reactor Mixed Oxide and Uranium Oxide (PWR MOX/UO2) Core Transient Benchmark case was chosen because it has been used widely to validate neutronic codes. The cross-section data in this study will be generated by SRAC, Serpent, and HELIOS codes. The NODAL3 code will be used to calculate neutronic parameters from each cross-section. The neutronic parameters calculated by NODAL3 are the effective multiplication factor (keff), control rod worth, critical boron concentration, and power distribution under Hot Zero Power (HZP) conditions. The Power-Weighted Error (PWE) and Error-Weighted Error (EWE), as a measure of the relative error in fuel assembly power, are less than 5 %, indicating that the calculation is consistent with DeCART as a reference. The difference in calculated radial power peaking factor for all three cross-sections to reference data reaches 6.284 % (G-3), 8.438 % (G-3), and 10.998 % (C-7), respectively, for SRAC, Serpent, and HELIOS. The axial power distribution calculated by NODAL3 at the top and bottom of the reactor core has a relative error that peaked at 16.60 %, 13.86 %, and 10.20 %, respectively, for cross-sections provided by SRAC, Serpent, and HELIOS. Further improvements are needed for NODAL3 by applying various discontinuity factors to improve its performance.

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Section: Introductionmentioning

confidence: 99%

Effect of cross-sections data on calculated static neutronic parameters of PWR MOX/UO2 core transient benchmark case using NODAL3 code

Luthfi,

Pinem

2023

Acta Polytech

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“…These identified issues, therefore, reduced the feasibility of using 100% conventional MOX fuel in place of UO 2 fuel cycle in a reactor core. However, studies by Reda, Gomaa, Bashter, Amin 17 and Fetterman 18 have demonstrated the feasibility of replacing a certain percentage of AP1000 reactor core fuel with the conventional MOX, which showed relative good neutronic properties. But the core experienced reduction on the moderator and fuel temperature coefficients of reactivity being important safety parameters, in which the extent of reduction depends on the number of fuel rod/assembly replaced.…”

Section: Introductionmentioning

confidence: 99%

“…But the core experienced reduction on the moderator and fuel temperature coefficients of reactivity being important safety parameters, in which the extent of reduction depends on the number of fuel rod/assembly replaced. Moreover, the replacement of UO 2 fuel rods/assemblies with the conventional MOX fuel still produced large‐scale high‐level radioactive isotopes, retaining strong proliferation risk and waste challenges 17,19 . These issues associated with the conventional MOX were found to be addressed by the thorium mixed oxide fuel 19 …”

Section: Introductionmentioning

confidence: 99%

Burn‐up calculation of the neutronic and safety parameters of thorium‐uranium mixed oxide fuel cycle in a Westinghouse small modular reactor

et al. 2020

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Summary Thorium fuel is presently a globally known future nuclear fuel alternative, having good neutronic, physical and chemical properties in addition to its spent nuclear fuel characteristic proliferation resistance. This research focused on the neutronic and safety parameters of thorium‐uranium mixed oxide fuel cycle, utilising three fissile enrichment zones, a departure from the conventional single enrichment. The aim was to determine the range of three fissile zones adequate for thorium‐uranium fuel cycle; investigating the performance efficiency of the fuel neutronic and inherent safety parameters in response to temperature differentials, which determines the viability of the fuel and core composition. Use was made of the MCNPX 2.7 code integrated with the CINDER90 fuel depletion code for steady‐state and burn‐up calculations. The keff, moderator temperature coefficient (MTC) and fuel temperature coefficient (FTC) of reactivity are affected by the range of fissile enrichment and fuel temperature which decreased with their respective increases. The MTC for all the moderator temperatures was within 0 to −40 pcm/K design value for UO2 fuel. Similarly, the FTC was within −3.5 to −1 pcm/K design value for all the fuel temperatures except after 2000 days, where a positive reactivity feedback was introduced. At ~86 MWd/kgHM single discharge burn‐up, the result shows that ~90% of the initial fissile load was utilised for energy production at the normal reactor operating temperature (600 K) with a slight reduction at higher fuel temperature. The total fissile inventory ratio (FIR), 233U/kg‐232Th and 239Pu/kg‐238U inventory ratios were significantly large and increased with burn‐up. It is remarkable that the FIR and the 233U/kg‐232Th inventory ratio did not reach conversion equilibrium until exit burn‐up. The large percentage fuel utilisation supports the advantage of fissile enrichment zoning in a thermal nuclear reactor core, making the chosen novel three fissile enrichment zones for thorium‐uranium fuel cycle reliable.

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“…The main advantages of recycling plutonium to make a MOX fuels are reducing the amount of enriched uranium and reducing radioactive waste generated from spent nuclear fuel [1]. For this reason, research is still being carried out regarding the use of MOX fuel in PWR reactors [2][3] [4]. Most light water-cooled (LWR) reactors have been licensed to use MOX fuels at fraction up to 30% or more on the reactor core [5] [6].…”

Section: Introduction *mentioning

confidence: 99%

Calculation of 2-Dimensional PWR Mox/Uo2 Core Benchmark Oecd Nea 6048 With Srac Code

Luthfi

Pinem

2020

Tri Dasa Mega

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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

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

Cited by 7 publications

References 6 publications

Effect of cross-sections data on calculated static neutronic parameters of PWR MOX/UO2 core transient benchmark case using NODAL3 code

Effect of cross-sections data on calculated static neutronic parameters of PWR MOX/UO2 core transient benchmark case using NODAL3 code

Burn‐up calculation of the neutronic and safety parameters of thorium‐uranium mixed oxide fuel cycle in a Westinghouse small modular reactor

Calculation of 2-Dimensional PWR Mox/Uo2 Core Benchmark Oecd Nea 6048 With Srac Code

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