Preliminary Study on Minor Actinide Incineration in RSG-GAS without Isotope Separation

Minor actinides (MA) resulted from nuclear power plants is often considered as nuisance in spent fuel management due to its considerably long half-life. One of available strategies to deal with MA is to incinerate it, in order to reduce its radioactivity. This paper presents a study on MA incineration in RSG-GAS research reactor. Unlike previous study, this work did not separate the MA into individual isotopes, but incinerated as a whole. ORIGEN2.1 code is employed to calculate MA incineration within RSG-GAS core. MA composition used in this study consists of Np, Am, and Cm isotopes. The Central Irradiation Position (CIP) of RSG-GAS is loaded by 6 kg of MA and irradiated for two years. The result shows that about 1 kg of MA were incinerated after two years of irradiation, or 18,87% of the initial concentration. However, the increase of Cm-242 isotope, along with newly-formed Pu isotopes, were found to be significantly increasing short-term radioactivity compared to un-irradiated MA. Thus, two years-worth of MA incineration cannot be considered as effective, and other strategies must be pursued.

A Laser Irradiation Method for Controlling Pieris rapae Larvae

At present, chemical pesticides remain the main approach for controlling Pieris rapae (L.) (Lepidoptera: Pieridae). This research proposes a novel laser irradiation method for managing P. rapae larvae as an alternative to chemical control. The effectiveness of controlling larvae and the influencing factors of lasers were studied to estimate optimal parameter combinations. Tests using the antifeedant effect and mortality of the larvae as dependent variables showed that the laser power, irradiation area, laser opening time and irradiation position were positively correlated with the P. rapae controlling effect. The optimal parameters for each factor were the following: laser power, 7.5 W; irradiation area, 6.189 mm2; laser opening time, 1.177 s; and irradiation position, middle of the abdomen. Based on these observations, a validation experiment was performed using the optimal combination of parameters, and the results showed that the antifeedant percentage of P. rapae larvae within 24 h posttreatment was 98.49%, whereas the mortality rate was 100%. The optimal parameter combination identified in the study was suitable for P. rapae larvae from the first- to fifth-instar stages, and a more effective controlling effect was observed with the younger larvae. These results can provide a theoretical basis for future pest control using laser pest-killing robots.

Determination of thermal neutron flux distribution at the rotary rack served for elemental concentration analysis using the k0-INAA method

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.

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

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.

DESIGN OF IRRADIATION FACILITIES AT CENTRAL IRRADIATION POSITION OF PLATE TYPE RESEARCH REACTOR BANDUNG

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

KARAKTERISASI LIMBAH DARI PRODUKSI RADIOISOTOP MOLIBDENUM-99

KARAKTERISASI LIMBAH DARI PRODUKSI RADIOISOTOP MOLIBDENUM-99. Radioisotop 99Mo diproduksi terutama sebagai radioisotop induk untuk memperoleh radioisotop tecnisium-99m (99mTc). Radioisotop 99mTc dipakai dalam kedokteran nuklir antara lain untuk diagnosis pada kelainan tulang, otak, thyroid, paru-paru, hati dan ginjal. Di Indonesia 99Mo diproduksi oleh PT. Industri Nuklir Indonesia (INUKI) dari target uranium yang dilekatkan kedalam dinding kapsul baja tahan karat untuk kemudian diiradiasi di Reaktor GA Siwabessy. Pengambilan 99Mo dalam target dilakukan dengan proses CINTICHEM. Pada proses CINTICHEM akan ditimbulkan beberapa jenis limbah yang salah satunya adalah Radioactive Fission Waste (RFW). Limbah ini memiliki paparan radiasi yang besar yang mengakibatkan karakterisasi limbah secara laboratorium sulit dilakukan. Pengelolaan limbah yang tepat memerlukan karakteristik limbah yang tepat pula. Oleh karena itu dalam penelitian ini dilakukan karakterisasi limbah RFW menggunakan program komputer ORIGEN 2.1 dengan data parameter input adalah data dari salah satu batch produksi 99Mo di PT INUKI yang berupa data target uranium diperkaya tinggi 92,7% yang dilekatkan pada kapsul baja tahan karat AISI 304L, iradiasi dilakukan pada posisi Centre Irradiation Position (CIP) dalam Reaktor Serbaguna GA Siwabessy dengan fluks netron termal: 1,12x1014 n/cm2detik dan iradiasi target dilakukan selama 96 jam. Seleksi radionuklida yang relevan terhadap metode pengelolaan limbah dilakukan berdasarkan pada waktu paro, tingkat kliren dan radiotoksisitas. Hasil penelitian menunjukkan bahwa sampai dengan waktu peluruhan 50 tahun, total konsentrasi aktivitas limbah 3,01x109 Bq/g dengan kandungan radionuklida produk aktivasi, aktinida dan anak luruhnya serta produk fisi. Selain itu limbah ini juga mengandung 235U yang masih cukup besar serta radionuklida umur paro panjang dengan tingkat toksisitas yang sangat tinggi. Berdasarkan pada Peraturan Pemerintah No.61 Tahun 2013 limbah ini diklasifikasikan sebagai limbah radioaktif tingkat sedang dan memerlukan pengelolaan dengan tingkat keselamatan yang tinggi.

ANALISIS OPTIMASI TARGET U-235 PENGKAYAAN RENDAH (LEU) UNTUK MENDUKUNG PRODUKSI 99Mo DAN TARGET BATU TOPAZ DI REAKTOR RSG-GAS

ANALISIS OPTIMASI TARGET U-235 PENGKAYAAN RENDAH (LEU) UNTUK MENDUKUNGPRODUKSI 99Mo DAN TARGET BATU TOPAZDI REAKTOR RSG-GAS. Peningkatan utilitas produksi radioisotop di reaktor RSG-GAS, seperti produksi99Mo dan target batu topaz perlu upaya optimasi iradiasi target U-235 pengkayaan rendah (LEU) di dalam teras reaktor RSGF-GAS. Terdapat 4 posisi IP (Irradiation Position) yang saat ini digunakan untuk iradiasi batu topaz secara rutin dari 8 posisi iradiasi yang tersedia di dalam teras reaktor. Oleh karena itu, iradiasi target yang lain hanya dapat dilakukan di posisi CIP (Central Irradiation Position). Penelitian ini bertujuan untuk menganalisis jumlah target LEU yang optimum di posisi CIP dan iradiasi batu topaz di IP RSG-GAS. Ruang lingkup penelitian ini meliputi pemantauan batasan keselamatan operasi faktor puncak daya radial maksimum, margin reaktivitas padam saat stuck rod dan perubahan fluks neutron di detektor un-balanced load. Pendekatan perhitungan menggunakan metode difusi neutron 2-dimensi, Batan-2DIFF dengan asumsi target LEU menggunakan proses electroplating. Hasil perhitungan menunjukkan jika 4 posisi CIP diisi penuh target LEUdengan massa optimal 14,4 g (1,2 g per kapsul) dan 36 g (3 g per kapsul) diperoleh 0,72% ∆k/k dan 1,12% ∆k/k jika dibandingkan dengan safety margin sebesar 2% ∆k/k maka seluruh parameter teras masih memenuhi syarat batas keselamatan operasi. Dengan demikian reaktor RSG-GAS aman dioperasikan untuk mengiradiasi targetLEU dengan massa yang optimal.

MEDAPP: Fission neutron beam for science, medicine, and industry

The instrument MEDAPP <strong>Med</strong>ical <strong>App</strong>lications), operated by the Technische Universität München, and the respective irradiation position are located at the world-wide unique fast neutron beam tube SR10 to which a uranium converter is attached. Thus, the instrument is operated with unmoderated fission neutrons and can be used for a broad variety of applications. For selected tasks, an alternative use with thermal neutrons is possible.