Studi Parameter Desain Teras Integral Pressurized Water Reactor Dengan Bahan Bakar Mixed Oxide Fuel  Menggunakan Program Srac

Diffusion approximation is an important approximation used to model a nuclear reactor core with a quite good accuracy and less computational cost. This approximation has been used widely around the globe for various kinds of nuclear reactors. This diffusion approximation is applied in a two-step method, a method combining a high fidelity transport code and low fidelity diffusion code. Meanwhile, innovations in the nuclear core model continue to make the nuclear reactor core safer, more robust, and smaller. The trend of creating smaller and more modular reactor core is emerging in the last ten years. These innovations will affect the core modeling system. Consequently, for smaller reactors, it is important to evaluate the capability of diffusion approximation if one wants to use a computationally cheaper method to model the reactor core. In this paper, neutron diffusion calculation for 160 MWth integral pressurized water reactor (IPWR) core was conducted using the PARCS nodal diffusion code employing the few-group spatially homogenized cross-sections generated by the Serpent Monte Carlo code. Due to its capability to model any reactor geometry in the high-resolution calculation, the results from Serpent were also used as a reference. Two important parameters are compared between PARCS and Serpent: effective neutron multiplication factor and core power distribution. For the full IPWR core model, a discrepancy of 564 pcm between PARCS and Serpent k eff was observed, while the radial power distribution had a maximum error of 4.71 %. It can be said, to some extent, that the diffusion approximation can be applied to IPWR core analysis. However, further improvement is indeed required if one wants more accurate results with low computational costs.

show abstract

“…The data of material and geometry used in the Serpent for Monte Carlo calculation is described in Table 1 [13,14].…”

Section: Methodsmentioning

confidence: 99%

Evaluating the Diffusion Approximation Capability on the Integral Pressurized Water Reactor (IPWR) Core Calculation

Ardiansyah

Oktavian²

2021

Atom Indo.

View full text Add to dashboard Cite

show abstract

“…The paper concludes that some property of the thorium-based fuel to the U-233 enrichment and the moderation ratio or fuel fraction [13]. The other research [14] about core design parametric study of integral pressurized water reactor (IPWR) with mixed oxide ceramic fuel using SRAC code system. The period of criticality of reactor core up to 1521 days with lowest CR is 0,622004 [14].…”

Section: Literature Review and Problem Statementmentioning

confidence: 99%

“…The other research [14] about core design parametric study of integral pressurized water reactor (IPWR) with mixed oxide ceramic fuel using SRAC code system. The period of criticality of reactor core up to 1521 days with lowest CR is 0,622004 [14]. The calculation of 2-Dimensional PWR MOX/UO 2 Core Benchmark OECD NEA 6048 with SRAC Code has been done [15].…”

Section: Literature Review and Problem Statementmentioning

confidence: 99%

“…This research can reach burn up time more than 20 years with excess reactivity less than 1 percent (∆k/k<1 %) both UN-PuN and ThN fuel. However, from the paper [8][9][10][11][12][13][14][15][16][17][18], all this suggests that it is advisable to conduct a study on comparing the optimization value of the neutronic calculation between UN and ThN fuel in PWR reactor (thermal reactor).…”

Section: Literature Review and Problem Statementmentioning

confidence: 99%

See 1 more Smart Citation

Comparison of thorium nitride and uranium nitride fuel on small modular pressurized water reactor in neutronic analysis using SRAC code

Syarifah¹,

Aula²,

Ardianingrum³

et al. 2022

EEJET

View full text Add to dashboard Cite

Comparison of thorium nitride (ThN) and uranium nitride (UN) fuel on small modular PWR in neutronic analysis has been carried out. PWR in module is one type of reactor that can be utilized because of its small size so that it can be placed on demand. Neutronic calculations were performed using SRAC version 2006, the data library using JENDL 4.0. The first calculation was fuel pin (PIJ) calculation with hexagonal fuel pin cell type. And the second calculation was reactor core (CITATION) calculation using homogeneous and heterogeneous core configurations. ThN and UN fuels use heterogeneous configurations with 3 fuel variations. The reactor geometry was used in two fuels are the same, with diameter and height active core was 300 cm and 100 cm. In this research, Np-237 was added as a minor actinide in the UN fuel to reduce the amount of Np-237 in the world and also reduce the k-eff value. For ThN fuel, Pa-231 also added in the fuel to reduce the k-eff value. The optimum configuration of UN fuel reached when used heterogeneous core configuration case four with percentage of U-235 in F1=5.5 %, F2=7 % and F3=8.5 % also with the addition of Np-237 0.2 % and fuel fraction 56 %. It has a maximum excess reactivity value 12.56 % %∆k/k. And then, the optimum configuration of ThN fuel reached when used heterogeneous core configuration case three with percentage of U-233 in F1=2 %, F2=4 % and F3=6 % with the addition of Pa-231 0.5 % and fuel fraction 53 %. It has a maximum excess reactivity value 7.67 % %∆k/k. The comparison of optimum design of UN and ThN fuel shows that the ThN fuel has the k-eff value closer to critical than UN fuel. Therefore, in this study, ThN fuel is more suitable for use in PWR reactors because it has a small excess value and can operate for 10 years without refueling

show abstract

Perhitungan Burn Up pada Reaktor HCLWR Model Geometri Heksagonal Dua Dimensi BahanBakar Uranium menggunakan SRAC COREBN

Ardanti

Yulianti

Riyanto

et al. 2020

J. Energy Mater. Instrum. Technol.

View full text Add to dashboard Cite

This study about burn up calculations on the HCLWR reactor with two dimensional hexagonal geometric models of uranium using SRAC COREBN has been carried. This study used uranium-235 and uranium-238 as the fuels, light water as coolant and moderator. The calculation was done with computational simulation of COREBN. The calculation aims to produce a conversion ratio more than 1 and a critical multiplication factor (keff). The research used fuels enrichment, burn up period, burn up step, linier power and volume fraction parameters. The percentage of enrichment from U1 and U3 were 3% and U2 was 2,5%. The volume fraction of each material was 52% fuels, 12% cladding, and 36% coolant, the reactor operating period for 548 days with 6 step of burn up then the reactorpower is 2400 MW and the linear poweris 1,792722 MW/cm. The calculation of conversion ratio was 1,01233 and keff was 0,901.

show abstract

Studi Parameter Desain Teras Integral Pressurized Water Reactor Dengan Bahan Bakar Mixed Oxide Fuel Menggunakan Program Srac

Cited by 3 publications

References 0 publications

Evaluating the Diffusion Approximation Capability on the Integral Pressurized Water Reactor (IPWR) Core Calculation

Evaluating the Diffusion Approximation Capability on the Integral Pressurized Water Reactor (IPWR) Core Calculation

Comparison of thorium nitride and uranium nitride fuel on small modular pressurized water reactor in neutronic analysis using SRAC code

Perhitungan Burn Up pada Reaktor HCLWR Model Geometri Heksagonal Dua Dimensi BahanBakar Uranium menggunakan SRAC COREBN

Contact Info

Product

Resources

About