Kinetics parameters evaluation on the first core of the RSG GAS (MPR-30) using continuous energy Monte Carlo method

Liem, Peng Hong; Surbakti, TS Tukiran; Hartanto, Donny

doi:10.1016/j.pnucene.2018.08.014

Cited by 23 publications

(16 citation statements)

References 11 publications

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“…However, it should be mentioned that the value of the k e f f at the full core configuration was not purely experimental data. The data recorded for this core configuration was given in unit of $ and then it was converted to k e f f using the computed total effective delayed neutron fraction (β e f f ) by Monte Carlo method [4] since there was no measured data available for β e f f . In addition, the k e f f Critical JDA01/E-9 (600 mm/284 mm) Critical JDA03/F-8 (600 mm/293 mm) Critical JDA05/C-5 (600 mm/288 mm) Critical JDA04/F-5 (600 mm/290 mm) Critical JDA07/D-4 (600 mm/282 mm) Critical…”

Section: Resultsmentioning

confidence: 99%

“…The results are then compared to the available experiment results for the effective neutron multiplication factor (k e f f )and the control rod worth. Meanwhile the neutron spectra, kinetics parameters, and the fission rate distribution of RSG GAS first core using ENDF/B-VIII.0 are also compared with the previous studies [3][4][5] which also have been performed by using MCNP6 in conjunction with previous version of nuclear data libraries such as ENDF/B-VII.1 [6] and JENDL-4.0 [7].…”

Section: Introductionmentioning

confidence: 99%

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Benchmarking the new ENDF/B-VIII.0 nuclear data library for the first core of Indonesian multipurpose research reactor (RSG GAS)

et al. 2020

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The Indonesian Multipurpose Research Reactor namely Reaktor Serba Guna G.A. Siwabessy (RSG GAS) is a 30 MWth (max.) pool-type reactor loaded with plate-type low-enriched uranium fuel, using light water as coolant and moderator, and beryllium as reflector. The benchmark of the 1st criticality core of RSG GAS using different nuclear data libraries such as JENDL-4.0, JENDL-3.3, ENDF/B-VII.0, and JEFF-3.1 have been performed in the previous work and compared with the experiment result. In this work, the newly released ENDF/B-VIII.0 neutron reaction and thermal neutron scattering libraries will be used and the important neu-tronics parameters such as multiplication factor, kinetics parameters, and fission reaction rate will be calculated using Monte Carlo code MCNP6.2 and compared against the previous work and the experiment result.

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Section: Resultsmentioning

confidence: 99%

Section: Introductionmentioning

confidence: 99%

Benchmarking the new ENDF/B-VIII.0 nuclear data library for the first core of Indonesian multipurpose research reactor (RSG GAS)

et al. 2020

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“…Di samping itu, paket program WIMSD-5B menggunakan data nuklir yang relatif baru yaitu ENDF/B-VII.1 [13,14]. Peningkatan akurasi perhitungan dengan ENDF/B-VII.1 telah dipublikasikan dalam penelitian terkait [3,9,15,16].…”

Section: Pendahuluanunclassified

Analisis Akurasi Paket Program Wimsd-5b/Citation Dalam Perhitungan Kritikalitas Reaktor Msre Accuration Analysis of the Wimsd-5b/Citation Codes on the Criticality Calculation of the Msre Reactor

Permana

Tamaza²,

Sid'qon³

2021

gnd

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Paket program WIMSD-5B/CITATION telah digunakan sebagai perangkat analitik kekritisan berbagai jenis reaktor dan memberikan hasil yang memuaskan. Meskipun demikian, paket program ini belum pernah digunakan untuk menganalisis kekritisan reaktor jenis MSR (Molten Salt Reactor). Dalam penelitian ini program WIMSD-5B/CITATION dianalisis akurasinya untuk menghitung kritikalitas reaktor MSRE (Molten Salt Reactor Experiment), yaitu reaktor MSR yang pernah dioperasikan sebagai fasilitas eksperimen. Tujuan penelitian ini adalah untuk mengetahui akurasi WIMSD-5B/CITATION untuk menghitung kritikalitas bahan bakar Tipe C di reaktor MSRE. Kritikalitas teras reaktor MSRE dihitung dengan 6 kelompok energi neutron dengan model geometri R-Z. Hasil perhitungan menunjukkan model sel yang menempatkan celah di tengah grafit (Model 1) lebih baik dibanding dengan sel yang menempatkan celah bahan bakar di luar grafit (Model 2). Namun demikian perbedaan relatif antara hasil perhitungan dan n hasil eksperimen masih tinggi yaitu 7,23%. Akurasi perhitungan kritikalitas didominasi oleh faktor model geometri sel dan teras. Pengaruh signifikan juga bersumber dari data jumlah void dan komposisi pengotor Li-6. Hasil penelitian juga menunjukkan bahwa fluks neutron dan faktor puncak daya radial di teras MSRE sangat sensitif terhadap model sel bahan bakar.

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“…The part of the control rod that can interact with the fission reaction can be adjusted so that the value of the effective folding factor can be properly adjusted to keep the reactor critical. The control rod system can also be used to increase the power or turn off the reactor state, especially as a feature to shut down the reactor in an emergency manner by plugging the rod completely into the reactor core [4]. Each control rod device has a specific reactivity value, indicating its ability to absorb both high and low energy neutrons.…”

Section: Introduction mentioning

confidence: 99%

Analyses of Neutron Absorber Materials on the Safety Parameters in the RSG-Gas Reactor

Suparlina

Surbakti²,

Pinem³

et al. 2021

Tri Dasa Mega

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Shutdown system in RSG-GAS reactor is using neutron absorber. There are 3 kinds of absorber material in research reactors including Ag-In-Cd alloy, B4C, and Hf. In this works, analyses of different neutron absorbers on the main safety core parameters in the RSG-GAS research reactor are selected for analyses. Their integral effects on the main neutronic core parameters important to safety issues are investigated. These parameters are core excess reactivity, shutdown margin, total reactivity worth of control rods, PPF and neutron flux . The RSG-GAS core silicide fuel is selected as the case study to verify calculations. A three-dimensional, four-group diffusion model is selected for core calculations. The well-known WIMSD-5B and Batan-3DIFF reactor codes are used to carry out these calculations. It is found that the largest shutdown margin is gained using the B4C; also the lowest PPF is gained using the Hf material. The maximum point power densities belong to the inside fuel regions surrounding the CIP (centre irradiation position), surrounded by control fuel elements, and the peripheral fuel elements beside the berrylium reflector. The greatest and least fluctuation of the point power densities are gained by using B4C and Ag-In-Cd alloy, respectively.

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Kinetics parameters evaluation on the first core of the RSG GAS (MPR-30) using continuous energy Monte Carlo method

Cited by 23 publications

References 11 publications

Benchmarking the new ENDF/B-VIII.0 nuclear data library for the first core of Indonesian multipurpose research reactor (RSG GAS)

Benchmarking the new ENDF/B-VIII.0 nuclear data library for the first core of Indonesian multipurpose research reactor (RSG GAS)

Analisis Akurasi Paket Program Wimsd-5b/Citation Dalam Perhitungan Kritikalitas Reaktor Msre Accuration Analysis of the Wimsd-5b/Citation Codes on the Criticality Calculation of the Msre Reactor

Analyses of Neutron Absorber Materials on the Safety Parameters in the RSG-Gas Reactor

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