Compendium of Material Composition Data for Radiation Transport Modeling

Williams, Ralph G.; Gesh, Christopher J.; Pagh, Richard T.

doi:10.2172/1023125

Cited by 263 publications

(142 citation statements)

References 6 publications

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“…Gambar 1. Ilustrasi bahan bakar yang digunakan pada reaktor RDE [23]. Untuk komposisi material pada perisai biologis dari beton setebal 100 cm yang digunakan dalam perhitungan dengan memanfaatkan data referensi beton dari Los Alamos National Laboratory (LANL) -USA, seperti di sajikan dalam Tabel 2.…”

Section: Deskripsi Bahan Bakar Pebbleunclassified

Analisis Kuat Sumber Neutron Dan Perhitungan Laju Dosis Neutron Teras Awal Rde

Suwoto

Adrial

Zuhair

2017

urania

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ABSTRAK ANALISIS KUAT SUMBER NEUTRON DAN PERHITUNGAN LAJU DOSIS NEUTRON TERAS AWAL RDE.Teras reaktor RDE (Reaktor Daya Eksperimental) berbentuk silinder non anular, berbahan bakar kernel partikel berlapis TRISO yang berbentuk bola (pebble) dan berpendingin gas helium. Desain teras reaktor ini sangat aman karena mengadopsi teknologi reaktor temperatur tinggi HTGR (High Temperature Gas-cooled Reactor) dengan keselamatan inheren pasif. Temperatur keluaran gas helium teras reaktor dirancang pada kisaran 700 C dengan temperatur masukan sekitar 250 C. Di samping menghasilkan listrik, reaktor didisain menghasilkan panas temperatur tinggi yang dapat digunakan untuk keperluan kogenerasi lainnya (penelitian panas proses lainnya). Bahan bakar reaktor berbentuk bola yang berisikan kernel partikel berlapis TRISO yang berupa uranium dioksida (UO2) berpengkayaan 17 % ( 235 U). Lapisan TRISO terdiri 4 lapisan yaitu lapisan karbon penyangga berpori, lapisan karbon pirolitik bagian dalam (IPyC, Inner Pyrolitic Carbon), lapisan Silikon Karbida (SiC) dan lapisan pirolitik karbon bagian luar (OPyC, Outer Pyrolitic Carbon). Analisis kuat sumber dan perhitungan awal laju dosis neutron pada RDE dilakukan menggunakan program Monte Carlo MCNP5v1.2. Pemodelan heterogenitas ganda bahan bakar dilakukan pada bahan bakar kernel partikel berlapis TRISO dan pada bahan bakar bola pada teras reaktor. Pemanfaatan program EGS99304 untuk menentukan jumlah struktur group energi yaitu 640 (SAND-II group structure) yang digunakan dalam perhitungan spektrum neutron pada reaktor. Teras reaktor dibagi dalam 100 zona (10 arah radial dan 10 arah aksial). Analisis hasil perhitungan menunjukkan bahwa kuat sumber neutron reaktor sebesar 8,47027X10 17 neutron/detik. Distribusi laju dosis neutron ditentukan menggunakan faktor konversi fluks neutron ke dalam dosis neutron dari International Commission on Radiological Protection (ICRP) dan National Council on Radiation Protection and Measu-rements (NCRP). Hasil perhitungan awal laju dosis neutron dengan faktor konversi ICRP-21 dan NCRP-38 untuk pekerja radiasi pada arah radial di perisai biologis sudah melemah memberikan nilai masing-masing sebesar 6,69915 µSv/jam dan 6,9964 µSv/jam pada posisi 215 cm dari pusat teras reaktor. Hasil analisis ini mengindikasikan bahwa pekerja radiasi aman dan terlindungi dari sumber radiasi neutron sesuai dengan persyaratan regulasi yang berlaku. Secara keseluruhan, dari hasil analisis perhitungan tersebut tampak bahwa model perisai radiasi dan perisai biologis telah memenuhi standar keselamatan radiasi yang dipersyaratkan. Hasil penelitian ini dapat digunakan untuk melengkapi data desain ketebalan perisai radiasi reaktor.Kata kunci: TRISO, pebble, RDE, laju dosis neutron, ICRP dan NCRP.

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Section: Deskripsi Bahan Bakar Pebbleunclassified

Analisis Kuat Sumber Neutron Dan Perhitungan Laju Dosis Neutron Teras Awal Rde

Suwoto

Adrial

Zuhair

2017

urania

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“…RayXpert proposes a database of materials, whose compositions correspond to the one given in the PNNL-15870 rev 1 [3]. The choice of a given material for a given volume automatically attributes that composition to the volume.…”

Section: Rayxpert Version 1340mentioning

confidence: 99%

“…In order to be consistent with the calculations made with RayXpert (see 2.1), the composition of materials was also extracted from the PNNL-15870 rev 1 [3].…”

Section: Mcnpx 270mentioning

confidence: 99%

Evaluation of RayXpert^®for shielding design of medical facilities

Derreumaux¹,

Vecchiola²,

Geoffray

et al. 2017

EPJ Web Conf.

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Abstract. In a context of growing demands for expert evaluation concerning medical, industrial and research facilities, the French Institute for radiation protection and nuclear safety (IRSN) considered necessary to acquire new software for efficient dimensioning calculations. The selected software is RayXpert ® . Before using this software in routine, exposure and transmission calculations for some basic configurations were validated. The validation was performed by the calculation of gamma dose constants and tenth value layers (TVL) for usual shielding materials and for radioisotopes most used in therapy . Calculated values were compared with results obtained using MCNPX as a reference code and with published values. The impact of different calculation parameters, such as the source emission rays considered for calculation and the use of biasing techniques, was evaluated.

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“…This system is comprised of a fixed number of fuel assemblies inside a basket and neutron shielding, called the MPC unit, which can be transferred between different types of container, and the HI-STORM over-pack itself, a steel structure designed to be filled with concrete on site that will serve as the shielding for the fuel assemblies during the interim storage. After developing the basic computational model of MPC and HI-STORM over-pack using MCNP code, a variety of concrete compositions appropriate for use in shielding were identified from databases, and then a set of four different types were selected for additional radiation transport simulations and comparison with basic model simulations 12 . Besides the basic comparison between the four composition types, separate studies to determine the effects of density changes of a composition, and of how additional decay time changing the spectrum of fuel source would influence shielding were also performed.…”

Section: Thesis Overviewmentioning

confidence: 99%

Uncertainty Quantification of Concrete Utilized in Dry Cask Storage

et al. 2015

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The objective of this thesis is to quantify the uncertainty in radiation dose rate estimates outside of a used fuel dry cask storage unit due to the parametric variability of concrete composition and density. This requires the selection of a limited number of concrete compositions from a standardized database and the development of a reference dry cask model, which can be used to estimate dose rate from neutrons and gamma rays. The majority of the dose in all cases, typically 70%, depended on gamma rays produced by the fission products. Density was the dominant factor in determining the dose rate, as expected. Composition variation while density was held fixed, however, indicated that the precise composition has a large effect on the dose rates produced by neutrons and iii gamma rays produced by neutron capture, on the order of 70% or higher for test cases, with only a moderate impact on the dose rate from gamma rays produced by other sources, on the order of 20% for test cases. Alterations to the gamma energy spectra produced by additional decay uniformly lowered the dose, and did not significantly change comparative concrete performance. Overall results indicate that, due to the lack of standardization of concrete poured on site, additional safety factors may be necessary due to variation of shielding effectiveness. Thanks also go to my friends and colleagues and the nuclear engineering department faculty and staff for making my time at Texas A&M University a great experience.

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Compendium of Material Composition Data for Radiation Transport Modeling

Cited by 263 publications

References 6 publications

Analisis Kuat Sumber Neutron Dan Perhitungan Laju Dosis Neutron Teras Awal Rde

Analisis Kuat Sumber Neutron Dan Perhitungan Laju Dosis Neutron Teras Awal Rde

Evaluation of RayXpert^®for shielding design of medical facilities

Uncertainty Quantification of Concrete Utilized in Dry Cask Storage

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Compendium of Material Composition Data for Radiation Transport Modeling

Cited by 263 publications

References 6 publications

Analisis Kuat Sumber Neutron Dan Perhitungan Laju Dosis Neutron Teras Awal Rde

Analisis Kuat Sumber Neutron Dan Perhitungan Laju Dosis Neutron Teras Awal Rde

Evaluation of RayXpert®for shielding design of medical facilities

Uncertainty Quantification of Concrete Utilized in Dry Cask Storage

Contact Info

Product

Resources

About

Evaluation of RayXpert^®for shielding design of medical facilities