scholarly journals The phenomenon of neutron flux during irradiation of the 4.5%UO2 PWR pin at the power ramp test facility RSG-GAS

2019 ◽  
Author(s):  
E. F. Octaviana ◽  
K. Azizul ◽  
Suharyana ◽  
Soeparmi
Keyword(s):  
Neutron Flux ◽  
Test Facility ◽  
Ramp Test

scholarly journals IRRADIATION CHARACTERISTIC OF NATURAL UO2 PIN PHWR TARGET AT PRTF OF RSG – GAS CORE

2017 ◽  
Vol 19 (2) ◽  
pp. 71
Author(s):  
Jati Susilo ◽  
Tagor Malem Sembiring ◽  
Winter Dewayatna
Keyword(s):  
Neutron Flux ◽  
Nuclear Power ◽  
Irradiation Time ◽  
Thermal Power ◽  
Analytical Tool ◽  
Test Facility ◽  
Nuclear Power Reactor ◽  
Ramp Test ◽  
The Core ◽  
Fuel Pin

The RSG-GAS reactor has a facility for irradiation of the fuel pin of nuclear power reactor, namely Power Ramp Test Facility (PRTF). The in-house fabrication PWR fuel pin has prepared for irradiations in the PRTF facility, currently, while the various enrichments of uranium are analyzed using the analytical tool. In the next step, it is planned to perform an irradiation of PHWR fuel pin sample of natural UO2 in the facility. Before irradiation in the core, it should be analyzed by using the analytical tool. The objectives of this paper are to optimize irradiation time based on the burn-up, the generated linear power and the neutron flux level at the target. The 3-dimension calculations have been carried out by using the CITATION code in the SRAC2006 code system. Since the coolant of the reactor is H2O, the effect of moderators in the pressurized tube, H2O and D2O, were analyzed, as well as pellet radius and moderator densities. The calculation results show that the higher linear power as irradiation time longer is occurred preferably in the D2O moderator than in H2O. For the D2O moderator, the higher pressure affects the lower density and longer irradiation time. The maximum irradiation time for natural UO2 fuel pin with the pressurized D2O moderator is about 9.5×104 h, with the linear power of 700 W/cm. During irradiation, neutronic parameters of the core such as excess reactivity and ppf show a very small change, still far below design value.Keywords:  PHWR, Neutron Flux, Thermal Power, PRTF, RSG-GAS KARAKTERISTIK IRADIASI TARGET PIN PHWR UO2 ALAM PADA PRTF TERAS RSG – GAS. Teras RSG-GAS dilengkapi dengan fasilitas untuk uji iradiasi bahan bakar nuklir atau disebut dengan Power Ramp Test Fasility (PRTF). Saat ini sedang dilpersiapkan untuk dilakukan uji sample pin bahan bakar PWR pada fasilitas PRTF. Analisis terhadap uji iradiasi sample pellet UO2 dengan berbagai pengkayaan telah dilakukan menggunakan paket program komputer. Dimasa yang akan datang, uji iradiasi pin bahan bakar PHWR UO2 alam juga sedang dalam perencanaan. Sebelum diiradiasi di dalam teras, maka terlebih dahulu harus dilakukan analisis dengan menggunakan paket program komputer. Tujuan dari penelitian ini adalah optimasi uji iradiasi pin bahan bakar UO2 alam sebagai fungsi waktu iradiasi berdasarkan burn-up, daya linier dan fluks neutron. Perhitungan teras RSG-GAS dilakukan dengan paket program SRAC2006 modul CITATION dalam bentuk geometri 3 dimensi. Analisis dilakukan terhadap pengaruh penggunaan jenis moderator pada tabung tekan iradiasi (H2O dan D2O), perubahan ukuran pelllet UO2 dan perubahan besarnya densitas moderator D2O. Dari analisis hasil perhitungan diketahui bahwa semakin lama waktu iradiasi akan menghasilkan daya termal yang semakin besar jika menggunakan moderator D2O dibandingkan H2O. Semakin tinggi tekanan atau semakin kecil densitas moderator, maka akan menghasilkan daya termal yang semakin besar seiring bertambah lamanya waktu iradiasi. Batas maksimal waktu iradiasi untuk pin bahan bakar UO2 alam dengan moderator D2O bertekanan adalah sekitar 9,5×104 jam, dengan batasan daya linier desain kemampuan peralatan, 700 W/cm. Selama iradiasi, nilai parameter neutronik teras reaktor seperti reaktivitas lebih dan ppf hanya menunjukkan perubahan yang sangat kecil, masih jauh dibawah batas yang ditetapkan dalam desain.Kata kunci: PHWR, Fluks Neutron, Daya Termal, PRTF, RSG-GAS

Test facility for a neutron flux spectrometer system based on the foil activation technique for neutronics experiments with the ITER TBM

2011 ◽  
Vol 86 (9-11) ◽  
pp. 2322-2325 ◽  
Author(s):  
A. Klix ◽  
A. Domula ◽  
U. Fischer ◽  
D. Gehre ◽  
P. Pereslavtsev ◽  
...  
Keyword(s):  
Neutron Flux ◽  
Test Facility ◽  
Activation Technique ◽  
Foil Activation ◽  
Spectrometer System

scholarly journals Perhitungan Perubahan Reaktivitas Pada Reaktor Serba Guna G.A. Siwabessy Akibat Pengoperasian Power Ramp Test Facility

2019 ◽  
Vol 4 ◽  
pp. 121
Author(s):  
A Suparmi ◽  
Tuti Dwi Setyaningsih ◽  
Suharyana Suharyana ◽  
Fuad Anwar ◽  
Riyatun Riyatun
Keyword(s):  
Reactor Core ◽  
Test Facility ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Water Reactor ◽  
Ramp Test ◽  
Fuel Pin ◽  
Safe Limit ◽  
Reactor Operating ◽  
Test Fuel

<p><strong>Abstract: </strong>Power Ramp Test Facility (PRTF) is one of the irradiation facility contained in the Multipurpose Reactor GA Siwabessy. This facility is used to test the reactor fuel element pin-type Pressurized Water Reactor. As a result of the entry of foreign bodies cause changes reactor conditions, one of which is expressed with the amount of reactivity to assess the safety of the reactor due to the operation PRTF. PRTF operation simulation and calculation is done using software neutronics MCNP6. Test UO2 fuel enriched assumed at 5% with constant power reactor operating at 15 MW and test fuel pin placed on PRTF within 0, 20, 40, 60, 80, 100, 120, and 140 mm from the centre of the reactor core. Change of reactivity values required in order to secure the reactor, maximal value is 0,5%<em></em>.  The calculation were obtained at each position is (<em></em><em></em>;  <em></em>;  <em></em>; <em></em>;<em></em>; <em></em>; <em></em>; <em></em>). Change of reactivity values smaller than the safe limit. Therefore, the study of reactivity changes PRTF operation to test fuel pin is secure.</p><p><strong>Abstrak: </strong>Power Ramp Test Facility (PRTF) merupakan salah satu fasilitas iradiasi yang terdapat pada Reaktor Serba Guna G.A. Siwabessy. Fasilitas ini digunakan untuk menguji pin elemen bahan bakar reaktor tipe Pressurized Water Reactor. Akibat dari masuknya benda asing menyebabkan perubahan kondisi reaktor, salah satunya dinyatakan dengan besaran reaktivitas untuk mengkaji keselamatan reaktor akibat pengoperasian PRTF. Simulasi pengoperasian PRTF dan perhitungan netronik dilakukan menggunakan perangkat lunak MCNP6. Bahan bakar uji UO2 diasumsikan diperkaya sebesar 5% dengan daya operasi reaktor konstan sebesar 15 MW. Pin bahan bakar uji diletakkan pada PRTF berjarak 0, 20, 40, 60, 80, 100, 120, dan 140 mm dari arah pusat teras reaktor. Nilai perubahan reaktivitas yang dipersyaratkan agar reaktor aman adalah , sedangkan nilai perubahan reaktivitas dari penelitian pada masing-masing posisi dari pusat reactor adalah (;  ;  ; ;; ; ; ) . Nilai perubahan reaktivitas akibat masuknya pin bahan bakar di PRTF mempunyai nilai perubahan reaktivitas 1/10 kali lebih kecil daripada batas aman. Oleh karena itu, ditinjau dari kajian  nilai perubahan reaktivitas maka pengoperasian PRTF untuk uji pin bahan bakar adalah aman.</p>

Safety Analysis of Indonesia's Multipurpose Reactor During Irradiation of Natural UO2 Pin Target at Power Ramp Test Facility

2020 ◽  
Author(s):  
Octaviana Erawati Fadli ◽  
Azizul Khakim ◽  
Suharyana ◽  
A Suparmi
Keyword(s):  
Heat Rate ◽  
Reactor Power ◽  
Test Facility ◽  
Original Position ◽  
Target Movement ◽  
Ramp Test ◽  
Maximum Value ◽  
Safety Parameters ◽  
Reactivity Insertion ◽  
Safety Limit

Abstract In order to support the utilization of Power Ramp Test Facility (PRTF) at RSG-GAS reactor, the investigation of safety parameters both on the target and the reactor during the irradiation of natural UO2 has been conducted. The PARET/ANL and MCNP6 codes are integrated for thermal-hydraulic and neutronic analysis under the reactivity insertion phenomenon. In the MCNP6 code, the reactivity insertion and Linear Heat Rate (LHR) are analyzed to confirm that the maximum value is still below the safety limits. In addition, the code computes kinetic parameters to provide inputs for the PARET/ANL code to evaluate the reactor response due to the target movement at the PRTF. The target moves from the farthest position to the nearest one traveling 440 mm, and then moves back to the original position. The result of the irradiation target simulation is satisfying, both maximum reactivity insertion and LHR values do not exceed the safety limit allowed. The pin target induces maximum reactivity insertion of 0.023%dk/k and produces the maximum LHR of 137.27 watt/cm and 274.54 watt/cm; when the reactor operates at 15 MW and 30 MW, respectively. In addition, the PARET/ANL simulation indicates that the reactor power oscillates due to the target movement. In order to avoid the reactor scram, the recommended pin target maximum velocities are 170 mm/s and 80 mm/s for reactor power of 15 MW and 30 MW, respectively.

scholarly journals Development of a regenerated Beryllium target and a thermal test facility for Compact Accelerator-based Neutron Sources

2020 ◽  
Vol 231 ◽  
pp. 03003
Author(s):  
Jean-François Muraz ◽  
Daniel Santos ◽  
Véronique Ghetta ◽  
Julien Giraud ◽  
Julien Marpaud ◽  
...  
Keyword(s):  
Neutron Flux ◽  
Deuteron Beam ◽  
Test Facility ◽  
Tumour Site ◽  
Original Solution ◽  
Physical Constraints ◽  
Thermal Test ◽  
Energy Proton ◽  
Beryllium Target ◽  
Target Design

Recently, the possibility to use compact accelerators coupled to high current ion sources for the production of intense low energy proton or deuteron beams has motivated many research laboratories to develop accelerator based neutrons sources for several purposes, including Neutron Capture Therapy (NCT). The NCT needs a high flux, about 10 9 n.cm-2.s-1, of thermal neutrons (E<10 keV) at the tumour site. Up to now, the NCT required neutron flux was mainly delivered by nuclear reactors. However, the production of such neutron flux is now possible using proton or deuteron beams on specific targets able to stand a high pow er (~15- 30 kW) on a small area (~10 cm2). This specific target design, materials and supports, has to cope with extreme physical constraints . The LPSC team has conceived an original solution formed by a thin (8 μm) rotating beryllium target depos ited on a graphite wheel and coupled with a beryllium sputtering device for periodic 9Be layer restoration. By means of 9Be (d,n) 10B nuclear reaction, this target irradiated by a 10- -20 mA deuteron beam (1.45 MeV) should produce the required neutron flux. In order to validate the target design of the neutron flux production and the beryllium target thermal capabilities, we built a 30 cm diameter rotating Beryllium target prototype and a compact electron beam line able to deliver a power density of 3kW/cm2.

scholarly journals MODEL SIMULATION OF GEOMETRY AND STRESS-STRAIN VARIATION OF BATAN FUEL PIN PROTOTYPE DURING IRRADIATION TEST IN RSG-GAS REACTOR

2015 ◽  
Vol 21 (1) ◽  
Author(s):  
Suwardi Suwardi ◽  
Winter Dewayatna ◽  
Sungkono Sungkono ◽  
Ridwan Ridwan ◽  
M Rifai
Keyword(s):  
Thermal Expansion ◽  
Model Simulation ◽  
Test Facility ◽  
Strain Variation ◽  
Stress Strain ◽  
Ramp Test ◽  
Irradiation Test ◽  
Fuel Pin ◽  
Computation Data ◽  
Fuel Technology

MODEL SIMULATION OF GEOMETRY AND STRESS-STRAIN VARIATION OF BATAN FUEL PIN PROTOTYPE DURING IRRADIATION TEST IN RSG-GAS REACTOR*). The first short fuel pin containing natural UO2 pellet in Zry4 cladding has been prepared at the CNFT (Center for Nuclear Fuel Technology) then a ramp test will be performed. The present work is part of designing first irradiation experiments in the PRTF (Power Ramp Test Facility) of RSG-GAS 30 MW reactor. The thermal mechanic of the pin during irradiation has simulated. The geometry variation of pellet and cladding is modeled by taking into account different phenomena such as thermal expansion, densification, swelling by fission product, thermal creep and radiation growth. The cladding variation is modeled by thermal expansion, thermal and irradiation creeps. The material properties are modeled by MATPRO and standard numerical parameter of TRANSURANUS code. Results of irradiation simulation with 9 kW/m LHR indicates that pellet-clad contacts onset from 0.090 mm initial gaps after 806 d, when pellet radius expansion attain 0.015 mm while inner cladding creep-down 0.075 mm. A newer computation data show that the maximum measured LHR of n-UO2 pin in the PRTF 12.4 kW/m. The next simulation will be done with a higher LHR, up to ~ 25 kW/m.MODEL SIMULASI VARIASI GEOMETRI DAN STRESS-STRAIN DARI PROTOTIP BAHAN BAKAR PIN BATAN SELAMA UJI IRADIASI DI REAKTOR RSG-GAS. Pusat Teknologi Bahan Bakar Nuklir (PTBBN) telah menyiapkan tangkai (pin) bahan bakar pendek perdana yang berisi pelet UO2 alam dalam kelongsong paduan zircaloy untuk dilakukan uji iradiasi daya naik. Penelitian ini merupakan bagian dari perancangan percobaan iradiasi pertama di PRTF (Power Ramp Test Fasility) yang terpasang di reaktor serbaguna RSG-GAS berdaya 30 MW. Telah dilakukan pemodelan dan simulasi kinerja termal mekanikal pin selama iradiasi. Variasi geometri pelet dan kelongsong selama pengujian dimodelkan dengan memperhatikan fenomena ekspansi termal, densifikasi, bengkak oleh produk fisi, creep termal dan pertumbuhan iradiatif. Variasi sifat kelongsong dimodelkan oleh ekspansi termal, termal dan creep iradiatif. Sifat material dimodelkan dengan MATPRO serta parameter numerik standar kode TRANSURANUS. Hasil iradiasi simulasi dengan laju daya 9 kW/m, 75% data daya aksimal, menunjukkan bahwa awal kontak fisik pelet dengan kelongsong dari celah awal 0,09 mm terjadi setelah 806 hari, ketika ekspansi jejari pelet mencapai 0,015 mm sementara jejari kelongsong menyusut 0,075 mm. Data terbaru menunjukkan bahwa perhitungan maksimal dan pengukuran laju daya linear tangkai bahan bakar berisi UO2 alam di PRTF adalah 12,4 kW/m pada daya reactor 15 kW. Penelitian selanjutnya akan dilakukan dengan LHR lebih tinggi, sampai ~ 25 kW / m, bila daya reactor 30 MW. Keywords: iradiasi, pin bahan bakar, UO2 alam, geometri, tegangan-regangan.

Radiation Damage Studies with the HVEM

1972 ◽  
Vol 30 ◽  
pp. 680-681
Author(s):  
J. J. Laidler ◽  
B. Mastel
Keyword(s):  
Radiation Damage ◽  
Stainless Steels ◽  
Volume Expansion ◽  
Austenitic Stainless Steels ◽  
Test Facility ◽  
Fast Breeder Reactors ◽  
Breeder Reactors ◽  
Under Construction ◽  
Development Laboratory ◽  
Insight Into

One of the major materials problems encountered in the development of fast breeder reactors for commercial power generation is the phenomenon of swelling in core structural components and fuel cladding. This volume expansion, which is due to the retention of lattice vacancies by agglomeration into large polyhedral clusters (voids), may amount to ten percent or greater at goal fluences in some austenitic stainless steels. From a design standpoint, this is an undesirable situation, and it is necessary to obtain experimental confirmation that such excessive volume expansion will not occur in materials selected for core applications in the Fast Flux Test Facility, the prototypic LMFBR now under construction at the Hanford Engineering Development Laboratory (HEDL). The HEDL JEM-1000 1 MeV electron microscope is being used to provide an insight into trends of radiation damage accumulation in stainless steels, since it is possible to produce atom displacements at an accelerated rate with 1 MeV electrons, while the specimen is under continuous observation.

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