scholarly journals DESIGN OF NEUTRONIC PARAMETERS OF MTR REACTOR USING WIMSD-5B/BATAN-FUEL CODES

2020 ◽  
Vol 5 (3) ◽  
pp. 239-248
Author(s):  
Tukiran Surbakti ◽  
Surian Pinem ◽  
Lily Suparlina

BATAN has three aging research reactors, so it is necessary to design a new, more modern MTR type reactor using high-density, low enrichment uranium molybdenum fuel. The thermal neutron flux at the irradiation position is an important concern in the design of research reactors. This analysis is performed using standard computer codes WIMSD-5B and Batan-FUEL. The purpose of this study is to analyze the effect of the core configuration with safety control rods and neutronic parameters using the diffusion method calculation. The reactor core consists of 16 fuel elements and four control rods placed in the 5 x 5 position of the grid plate and is loaded the reflector elements outside the core. The cycle length is also a concern, not less than 20 days, and the reactor can be operated safely with a power of 50 MW. The calculation results show that for the highest fuel loading, which is 450 grams of U7Mo/Al fuel with D2O as a reflector, it will provide the lowest thermal neutron flux at the center of the core irradiation position, namely 1.0 x1015 n/cm2s. The core fuel cycle length will be up to 39 days, meeting the expected acceptance and safety criteria.

2021 ◽  
Vol 10 (1) ◽  
pp. 11-20
Author(s):  
Tho Nguyen Thi ◽  
Anh Tran Tuan ◽  
Cuong Trinh Van ◽  
Doanh Ho Van ◽  
Duong Tran Quoc ◽  
...  

The accuracy of elements concentration determination using the k0-standardization method directly depends on irradiation and measurement parameters including Non-1/E epithermal neutron flux distribution shape α (ϕ epi ≈1/E1+α ) , thermal-to-epithermal neutron flux ratio f, efficiency ε, peak area… In the case of the irradiation position at the rotary rack of the Dalat Nuclear Research Reactor (DNRR), the difference of thermal neutron flux between the bottom (3.54x1012 n.cm-2.s-1) and the top (1.93x1012 n.cm-2.s-1) of the 15 cm aluminum container is up to 45%. Therefore, it is necessary to accurately determine above-mentioned parameters in the sample irradiation position. The present paper deals with the determination of the distribution of thermal neutron flux along the sample irradiation container by using 0.1% Au–Al wire activation technique. The thermal neutron flux was then used to calculate the concentration of elements in the Standard Reference Material 2711a and SMELS type III using k0-INAA method at different positions in the container. The obtained results with the neutron flux correction were found to be in good agreement with the certified values. In conclusion, the proposed technique can be applied for activation analyses without sandwiching flux monitors between samples during irradiations.


2018 ◽  
Vol 170 ◽  
pp. 04002 ◽  
Author(s):  
H. Carcreff ◽  
L. Salmon ◽  
V. Lepeltier ◽  
J.M. Guyot ◽  
E. Bouard

Nuclear heating inside an MTR reactor needs to be known in order to design and to run irradiation experiments which have to fulfill target temperature constraints. To improve the nuclear heating knowledge, an innovative calorimetric system CALMOS has been studied, manufactured and tested for the 70MWth OSIRIS reactor operated by CEA. This device is based on a mobile calorimetric probe which can be inserted in any in-core experimental location and can be moved axially from the bottom of the core to 1000 mm above the core mid-plane. Obtained results and advantages brought by the first CALMOS-1 equipment have been already presented. However, some difficulties appeared with this first version. A thermal limitation in cells did not allow to monitor nuclear heating up to the 70 MW nominal power, and some significant discrepancies were observed at high heating rates between results deduced from the calibration and those obtained by the “zero method”. Taking this feedback into account, the new CALMOS-2 calorimeter has been designed both for extending the heating range up to 13W.g-1 and for improving the “zero method” measurement thanks to the implementation of a 4-wires technique. In addition, the new calorimeter has been designed as a real operational measurement system, well suited to characterize and to follow the radiation field evolution throughout the reactor cycle. To meet this requirement, a programmable system associated with a specific software allows automatic complete cell mobility in the core, the data acquisition and the measurements processing. This paper presents the analysis of results collected during the 2015 comprehensive measurement campaign. The 4-wires technique was tested up to around a 4 W.g-1 heating level and allowed to quantify discrepancies between “zero” and calibration methods. Thermal neutron flux and nuclear heating measurements from CALMOS-1 and CALMOS-2 are compared. Thermal neutron flux distributions, obtained with the Self-Power Neutron Detector suited to the CALMOS-2 calorimetric probe, are compared with those obtained with current devices. This campaign allowed to highlight advantages brought by the human machine interface automation, which deeply refined the profiles definition. Finally, the decay of the reactor residual power after shutdown could be performed after shutdown, demonstrating the ability of such type of calorimeter to follow the heating level whatever the thermohydraulic conditions, forced or natural convection regimes.


2015 ◽  
Vol 37 ◽  
pp. 01024 ◽  
Author(s):  
Dmitry V. Gvozdyakov ◽  
Anton O. Tanishev ◽  
Svetlana A. Shvab ◽  
Vladimir N. Martyshev ◽  
Arian V. Kuzmin

2020 ◽  
Vol 22 (1) ◽  
pp. 1
Author(s):  
Epung Saepul Bahrum ◽  
Wawan Handiaga ◽  
Yudi Setiadi ◽  
Henky Wibowo ◽  
Prasetyo Basuki ◽  
...  

One of the results from Plate Type Research Reactor Bandung (PTRRB) research program is PTRRB core design. Previous study on PTRRB has not calculated neutron flux distribution at its central irradiation position (CIP). Distribution of neutron flux at CIP is of high importance especially in radioisotope production. In this study, CIP was modeled as a stack of four to five aluminum tubes (AT), each filled by four aluminum irradiation capsules (AIC). Considering AIC dimension and geometry, there are three possibilities of AT configuration. For irradiation sample, 1.45 gr of molybdenum (Mo) was put into AIC. Neutron flux distribution at Mo sample was calculated using TRIGA MCNP and MCNP software. The calculation was simulated at condition when fresh fuel is loaded into reactor core. Analyses of excess reactivity show that, after installing irradiation AT and Mo sample was put into each configuration, the excess reactivity is less than 10.9 %. The highest calculated thermal neutron flux at Mo sample is 5.08×1013 n/cm2.s at configuration 1. Meanwhile, the highest total neutron flux at Mo sample is located at capsule no. II and III. Thermal neutron flux profile is the same for all configurations. This result will be used as a basic data for PTRRB utilization.Keywords: Central Irradiation Position, Neutron Flux Distribution, MCNP, PTRRB


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