scholarly journals Thermal and epithermal neutron fluence rates in the irradiation facilities of the TRIGA IPR-R1 nuclear reactor

2010 ◽  
Vol 40 (1) ◽  
pp. 47-51 ◽  
Author(s):  
Dante Marco Zangirolami ◽  
Arno Heeren de Oliveira ◽  
Andréa Vidal Ferreira
Keyword(s):  
Nuclear Reactor ◽  
Epithermal Neutron ◽  
Neutron Fluence ◽  
Irradiation Facilities

scholarly journals ANALYSIS OF GAMMA HEATING AT TRIGA MARK REACTOR CORE BANDUNG USING PLATE TYPE FUEL

2016 ◽  
Vol 18 (3) ◽  
pp. 127 ◽  
Author(s):  
Setiyanto Setiyanto ◽  
Tukiran Surbakti
Keyword(s):  
Nuclear Reactor ◽  
Reactor Core ◽  
Empty Space ◽  
Computer Code ◽  
Safety Analysis ◽  
Design Calculation ◽  
Triga Reactor ◽  
Calculation Results ◽  
Fuel Elements ◽  
Irradiation Facilities

ABSTRACT In accordance with the discontinuation of TRIGA fuel element production by its producer, the operation of all TRIGA type reactor of at all over the word will be disturbed, as well as TRIGA reactor in Bandung. In order to support the continuous operation of Bandung TRIGA reactor, a study on utilization of fuel plate mode, as used at RSG-GAS reactor, to replace the cylindrical model has been done. Various assessments have been done, including core design calculation and its safety aspects. Based on the neutronic calculation, utilization of fuel plate shows that Bandung TRIGA reactor can be operated by 20 fuel elements only. Compared with the original core, the new reactor core configuration is smaller and it results in some empty space that can be used for in-core irradiation facilities. Due to the existing of in-core irradiation facilities, the gamma heating value became a new factor that should be evaluated for safety analysis. For this reason, the gamma heating for TRIGA Bandung reactor using fuel plate was calculated by Gamset computer code. The calculations based on linear attenuation equations, line sources and gamma propagation on space. Calculations were also done for reflector positions (Lazy Susan irradiation facilities) and central irradiation position (CIP), especially for any material samples. The calculation results show that gamma heating for CIP is significantly important (0,87 W/g), but very low value for Lazy Susan position (lest then 0,11 W/g). Based on this results, it can be concluded that the utilization of CIP as irradiation facilities need to consider of gamma heating as data for safety analysis report. Keywords: gamma heating, nuclear reactor, research reactor, reactor safety.   ABSTRAK Dengan dihentikannya produksi elemen bakar reaktor jenis Triga oleh produsen, maka semua reaktor TRIGA di dunia terganggu operasinya, termasuk juga reaktor TRIGA 2000 di Bandung. Untuk mendukung pengoperasian reaktor TRIGA Bandung, telah dilakukan kajian penggunaan bahan bakar jenis pelat seperti yang digunakan oleh RSG-GAS. Berbagai langkah analisis telah disiapkan, termasuk perhitungan desain teras, dan sistem keselamatannya. Penggunaan elemen bakar tipe pelat menghasilkkan reaktor dapat dioperasikan hanya dengan 20 elemen bakar. Dibandingkan teras aslinya, nampak bahwa teras baru menjadi lebih kecil dan kompak, rapat dayanya naik, tetapi menyisakan beberapa ruang kosong yang dimungkinkan untuk menempatkan fasilitas iradiasi di teras. Dengan adanya fasilitas iradiasi di dalam teras, maka pembangkitan panas gamma di teras menjadi faktor baru yang harus diperhatikan. Untuk alasan ini, telah dilakukan perhitungan pembangkitan panas gamma teras reaktor Triga 2000 Bandung mengunakan program Gamset. Perhitungan didasarkan pada persamaan atenuasi liner, sumber garis dan arah perambatan tiga dimensi. Selain panas gamma di teras, akan dihitung juga panas gamma di reflektor (Lazy Susan), dan di CIP untuk berbagai jenis bahan. Diperoleh hasil bahwa panas gamma di CIP cukup signifikan (0,87 w/g), tetapi di posisi Lazy Susan relatif kecil, rata-rata hanya 0,11 w/g. Dari hasil tersebut dapat disimpulkan bahwa penggunaan CIP untuk iradiasi perlu mempertimbangkan panas gamma dalam perhitungan LAK nya. Kata kunci: panas gamma, reaktor nuklir, reaktor penelitian, keselamatan reaktor 

scholarly journals JSI TRIGA neutron and gamma field characterization by TLD measurements

2020 ◽  
Vol 225 ◽  
pp. 04034
Author(s):  
Klemen Ambrožič ◽  
Klaudia Malik ◽  
Barkara Obryk ◽  
Luka Snoj
Keyword(s):  
Radiation Field ◽  
Radiation Effects ◽  
Nuclear Reactor ◽  
Measurement Techniques ◽  
Field Measurements ◽  
Reactor Irradiation ◽  
Fission Chamber ◽  
Start Up ◽  
Gamma Field ◽  
Irradiation Facilities

A well characterized radiation field inside a research nuclear reactor irradiation facilities enables precise qualification of radiation effects to the irradiated samples such as nuclear heating or changes in their electrical or material properties. To support the increased utilization of the JSI TRIGA reactor irradiation facilities in the past few years mainly on account of testing novel detector designs, electronic components and material samples, we are working on increasing the neutron and gamma field characterization accuracy using various modeling and measurement techniques. In this paper we present the dose field measurements using thermo-luminescent detectors (TLD’s) with different sensitivities neutron and gamma sensitivities, along with multiple ionization and fission chamber. Experiment was performed in several steps from reactor start-up, steady operation and a rapid shutdown, during which the ionization and fission chamber signals were acquires continuously, while the TLD’s were being irradiated at different stages during reactor operation and after shutdown, to also capture response to delayed neutron and gamma field. The results presented in this paper serve for validation of JSI designed JSIR2S code for delayed radiation field determination, initial results of its application on the JSI TRIGA TLD measurements will also be presented.

Reevaluation of Spectral Parameters and Neutron Fluxes in IC-7 Irradiation Channel of TRIGA MARK I IPR-R1 Research Nuclear Reactor

2021 ◽  
Author(s):  
Radojko Jacimovic ◽  
Maria Angela de Barros Correia Menezes
Keyword(s):  
Nuclear Reactor ◽  
Epithermal Neutron ◽  
Research Reactor ◽  
Nuclear Research ◽  
Irradiation Channel ◽  
Spectral Parameters ◽  
K0 Method ◽  
Nuclear Research Reactor ◽  
Research Nuclear Reactor ◽  
The Stability

Abstract The core configuration of the TRIGA MARK I IPR-R1 nuclear research reactor, Brazil, has been modified six times since the first criticality and the neutron fluxes have been determined using experimental and semi theoretical methodologies determining the neutron fluxes in different irradiation channels and devices, applying different procedures and materials. This reactor operates at 100 kW, however, after new configuration for 250 kW in 2001, the carousel no longer rotates during irradiations aiming at preserving the rotation mechanism. In 2003, the spectral parameters were determined experimentally by the "Cd-ratio for multi-monitor" in five specific channels aiming at the application of NAA k0-standardized method. The determinations were repeated applying the same procedure in 2016, 2018 and 2019. Values for thermal and epithermal neutron fluxes as well as f and a spectral parameters were determined. The experimental results for CRM BCR-320R were calculated by the k0-method of NAA, using the spectral parameters for irradiation channel IC-7 obtained in 2003, 2016, 2018 and 2019 and evaluated by En-score. The values showed that the differences in the results compared to those in 2003 were lower than 2.5%, inside the uncertainty of the method. It shows that the k0-method installed in CDTN is reliable and useful for various purposes. The results of the spectral parameter f presented small differences, in a period of 16 years, pointing out the stability of operation of the reactor TRIGA MARK I IPR-R1.

A new monitor for routine thermal and epithermal neutron fluence rate monitoring in k0 INAA

2008 ◽  
Vol 66 (12) ◽  
pp. 1964-1969 ◽  
Author(s):  
M.J.J. Koster-Ammerlaan ◽  
M.A. Bacchi ◽  
P. Bode ◽  
E.A. De Nadai Fernandes
Keyword(s):  
Epithermal Neutron ◽  
Neutron Fluence ◽  
Fluence Rate ◽  
Neutron Fluence Rate

Epithermal neutron activation analysis with a SLOWPOKE nuclear reactor

1981 ◽  
Vol 59 (10) ◽  
pp. 1470-1475 ◽  
Author(s):  
Donald Craig Stuart ◽  
Douglas Earl Ryan
Keyword(s):  
Neutron Activation Analysis ◽  
Activation Analysis ◽  
Neutron Activation ◽  
Nuclear Reactor ◽  
Epithermal Neutron ◽  
Instrumental Analysis ◽  
Epithermal Neutron Activation Analysis ◽  
Experimental Values ◽  
Selection Of

Epithermal neutron activation analysis is examined using a SLOWPOKE nuclear reactor. Data are given for 69 elements using cadmium and boron shielding. Cadmium ratios calculated from literature values of resonance integrals are compared with experimental values. All elements except cadmium are shielded more effectively by boron. Advantage factors for a selection of elements under cadmium, natural boron, and enriched boron shields are given. Boron shields are of particular practical value for rapid instrumental analysis.

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