scholarly journals Analisis Fraksi Volume Bahan Bakar Uranium Karbida Pada Reaktor Cepat Berpendingin Gas Menggunakan SRAC Code

2021 ◽  
Vol 3 (1) ◽  
pp. 13-18
Author(s):  
Ratna Dewi Syarifah ◽  
Nabil Nabhan MH ◽  
Zein Hanifah ◽  
Iklimatul Karomah ◽  
Ahmad Muzaki Mabruri
Keyword(s):  
Power Density ◽  
Volume Fraction ◽  
Japan Atomic Energy Agency ◽  
Energy Agency ◽  
Core Diameter ◽  
Homogeneous Core ◽  
Fuel Pin ◽  
Core Height ◽  
Data Libraries ◽  
Code Standard

Analysis of fuel volume fraction with uranium caride fuel in Gas Cooled Fast Reactor (GFR) with SRAC Code is has been done. The calculation used SRAC Code (Standard Reactor Analysis Code) which is developed by JAEA (Japan Atomic Energy Agency), and the data libraries nuclear used JENDL 4.0. There are two calculation has been used, fuel pin cell calculation (PIJ Calculation) and core calculation (CITATION Calculation). In core calculation, the leakage is calculated so the calculation more precise. The CITATION calculation use two type of core configuration, i.e. homogeneous core configuration and heterogeneous core configuration. The power density value of two type core configuration is quite difference. It is better use heterogeneous core configuration than homogeneous core configuration, because the power density of heterogeneous core configuration is flatter than the other. From the analysis of fuel volume fraction, when the volume fraction is increase, the k-eff value is increase. And the optimum design after has been analysis for fuel volume fraction, that is the fuel volume fraction is 49% with a heterogeneous core configuration of three types of fuel percentages, for Fuel1 9%, Fuel2 12% and Fuel3 15%. This reactor is cylindrical, has a core diameter of 240 cm and a core height of 100 cm.

Neutronic Analysis of Thorium Nitride (Th, U233)N Fuel for 500MWth Gas Cooled Fast Reactor (GFR) Long Life without Refueling

2017 ◽  
Vol 733 ◽  
pp. 47-50 ◽  
Author(s):  
Ratna Dewi Syarifah ◽  
Yacobus Yulianto ◽  
Zaki Su’ud ◽  
Khairul Basar ◽  
Dwi Irwanto
Keyword(s):  
Power Density ◽  
Fast Reactor ◽  
Average Power ◽  
Maximum Power ◽  
Maximum Power Density ◽  
Fuel Cladding ◽  
Power Density Distribution ◽  
Homogeneous Core ◽  
Long Life ◽  
Data Libraries

Neutronic analysis of Thorium Nitride (Th, U233)N fuel of 500MWth Gas Cooled Fast Reactor (GFR) has been done. In this study the neutronic analysis use SRAC2006 code both PIJ and CITATION calculation. The data libraries use JENDL 4.0. First calculation is survey parameter with U-233 enrichment variation. From the homogeneous core configuration calculation, when the enrichment of U-233 is 8.2%, the maximum k-eff value is 1,00819 with excess reactivity value 0,812%. The average power density is 63 Watt/cc and the maximum power density 100 Watt/cc. The heterogeneous core configuration calculation has been done to flattening the power of the reactor. The variation fuel of F1:F2:F3 = 7.8%:8%:8.8%. The fraction of fuel : cladding: coolant = 60%:10%:30%. The max k-eff value of heterogeneous core configuration is 1,01229 with excess reactivity value 1.21%. The average power density is 65 Watt/cc and the maximum power density 92 Watt/cc. The power density distribution of heterogeneous core configuration is flatter than homogeneous core configuration.

scholarly journals Sensitivity and Uncertainty of Criticality

2021 ◽  
pp. 215-243
Author(s):  
Masao Yamanaka
Keyword(s):  
Reactor Core ◽  
Nuclear Data ◽  
Japan Atomic Energy Agency ◽  
Reactor Physics ◽  
Energy Agency ◽  
Control Rod ◽  
The Core ◽  
Data Libraries ◽  
And Control ◽  
Nuclear Data Libraries

AbstractExcess reactivity and control rod worth are generally considered important reactor physics parameters for experimentally examining the neutron characteristics of criticality in a core, and for maintaining safe operation of the reactor core in terms of neutron multiplication in the core. For excess reactivity and control rod worth at KUCA, as well as at the Fast Critical Assembly in the Japan Atomic Energy Agency, special attention is given to analyzing the uncertainty induced by nuclear data libraries based on experimental data of criticality in representative cores (EE1 and E3 cores). Also, the effect of decreasing uncertainty on the accuracy of criticality is discussed in this study. At KUCA, experimental results are accumulated by measurements of excess reactivity and control rod worth. To evaluate the accuracy of experiments for benchmarks, the uncertainty originated from modeling of the core configuration should be discussed in addition to uncertainty induced by nuclear data, since the uncertainty from modeling has a potential to cover the eigenvalue bias more than uncertainty by nuclear data. Here, to investigate the uncertainty of criticality depending on the neutron spectrum of cores, it is very useful to analyze the reactivity of a large number of measurements in typical hard (EE1) and soft (E3) spectrum cores at KUCA.

Verification of Probabilistic Fracture Mechanics Analysis Code PASCAL Through Benchmark Analyses With FAVOR

2017 ◽  
Author(s):  
Yinsheng Li ◽  
Shumpei Uno ◽  
Jinya Katsuyama ◽  
Terry Dickson ◽  
Mark Kirk
Keyword(s):  
Fracture Mechanics ◽  
The United States ◽  
Pressure Vessels ◽  
Japan Atomic Energy Agency ◽  
Energy Agency ◽  
Benchmark Analysis ◽  
Probabilistic Fracture Mechanics ◽  
Pressurized Thermal Shock ◽  
Great Confidence ◽  
Reactor Pressure

A probabilistic fracture mechanics (PFM) analysis code called PASCAL has been developed by the Japan Atomic Energy Agency to evaluate failure frequencies of Japanese reactor pressure vessels (RPVs) during pressurized thermal shock (PTS) events based on Japanese data and Japanese methods published for or provided in Japanese codes and standards. To verify this code, benchmark analyses were carried out using the FAVOR code, which was developed in the United States and has been utilized in nuclear regulation. The results of these analyses confirmed with great confidence the applicability of PASCAL to failure probability and frequency evaluation of Japanese RPVs. An outline of PASCAL, the benchmark analysis conditions and analysis results are reported in this paper.

scholarly journals Recent activity of neutron diffractometers for biological crystallography at Japan Atomic Energy Agency

2010 ◽  
Vol 66 (a1) ◽  
pp. s121-s121
Author(s):  
Taro Tamada ◽  
Kazuo Kurihara ◽  
Takashi Ohhara ◽  
Nobuo Okazaki ◽  
Ryota Kuroki
Keyword(s):  
Atomic Energy ◽  
Japan Atomic Energy Agency ◽  
Energy Agency ◽  
Recent Activity

scholarly journals New features and improvements in the NEA nuclear data tool suite

2020 ◽  
Vol 239 ◽  
pp. 19003
Author(s):  
M. Fleming ◽  
I. Hill ◽  
J. Dyrda ◽  
L. Fiorito ◽  
N. Soppera ◽  
...  
Keyword(s):  
Relational Databases ◽  
Uncertainty Propagation ◽  
Nuclear Data ◽  
Data Types ◽  
Reactor Physics ◽  
Energy Agency ◽  
Order Of Magnitude ◽  
Data Files ◽  
Improved Performance ◽  
Data Libraries

The OECD Nuclear Energy Agency (NEA) has developed and maintains several products that are used in the verification and validation of nuclear data, including the Java-based Nuclear Data Information System (JANIS) and the Nuclear Data Sensitivity Tool (NDaST). These integrate other collections of the NEA, including the International Handbooks of benchmark experiments on Criticality Safety and Reactor Physics (ICSBEP and IRPhEP) and their supporting relational databases (DICE and IDAT). Recent development of the JANIS, DICE and NDaST systems have resulted in the ability to perform uncertainty propagation utilising Legendre polynomial sensitivities, calculation of case-to-case covariances and correlations, use of spectrum weighting in perturbations, calculation of statistical results with suites of randomly sampled nuclear data files and new command-line interfaces to automate analyses and generate XML outputs. All of the most recent, major nuclear data libraries have been fully processed and incorporated, along with new visualisation features for covariances and sensitivities, an expanded set of reaction channel definitions, and new EXFOR data types defined by the NRDC. Optimisation of numerical methods has also improved performance, with over order-of-magnitude speed-up in the case of sensitivity-uncertainty calculations.

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