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.
Molybdenum-99 (99Mo) is a parent radioisotope of Technetium-99m (99mTc) widely used in nuclear diagnostics. The production of this radioisotope by PT. INUKI generated radioactive fission waste (RFW) that theoretically contains239Pu and235U, posing a nuclear proliferation risk. This paper discusses the determination of radionuclides inventory in the RFW and the proposed strategy for its management. The radionuclides inventory in the RFW was calculated using ORIGEN 2.1 code. The input parameters were obtained from one batch of 99Mo production using high enriched uranium in PT. INUKI. The result showed that the RFW contained activation products, actinides, and fission products, including239Pu and235U. This result was then used for consideration of the management of the RFW. The concentration of 235U was reduced by a down-blending method. The proposed strategy to further manage the down-blended RFW was converting it to U3O8 solid form, placed in a canister, and eventually stored in the interim storage for high-level waste located in The Radioactive Waste Technology Center.
The 10 MW HTR Indonesian Experimental Power Reactor (RDE reactor) is designed identical with the HTR-10 in China, conceptually. However, the review results showed that the spent fuel cask model which is used between two reactors is fully different, such as size and capacity. The proposed cask model in RDE reactor can hold 15 times more fuel pebbles than HTR-10 has. This research activities deal with the subcriticality analysis for the spent fuel cask of RDE reactor if using the HTR-10 cask model. The subcriticality condition is designed to meet the limit of safety value. The objective of this research is to determine the subcriticality value in the normal and accident events for the spent fuel cask when it is in the reactor building and the spent fuel cask room. All calculations were carried out by MCNP6.1 code. The selected external events are the water ingress (reactor room), water flood and the combination event of water flood and earthquake. The calculation results showed that the maximum value of keff (3σ) are 0.47510 and 0.19214 for the cask in the reactor building and in the spent fuel cask room, respectively. This value is far from the limit value of 0.95. The calculation results showed that the spent fuel cask are in the safe condition eventhough in the worst combination events, the cask is flooded and earthquake. The HTR-10 spent fuel cask can be proposed as an alternative for the RDE reactor to get an efficient reactor building.Keywords: spent pebble fuel element, HTGR, subcriticality, MCNP6.1, RDE reactor ANALISIS SUBKRITIKALITAS PENYIMPAN BAHAN BAKAR BEKAS MODEL CASK REAKTOR HTR-10 UNTUK REAKTOR DAYA EKSPERIMENTAL 10 MW TERMAL. Reaktor Daya Eksperimental (RDE) secara konseptual didesain identik dengan reaktor HTR-10 di Tiongkok. Meskipun demikian, terdapat perbedaan yang signifikan untuk desain konseptual cask penyimpan bahan bakar bekas di kedua reaktor seperti dimensi dan kapasitas. Kegiatan penelitian ini berkaitan dengan analisis subkritikalitas cask penyimpan elemen bahan bakar bekas tipe pebble di RDE jika menggunakan model cask yang dipakai di HTR-10. Kondisi sub-kritikalitas didesain memenuhi nilai batas keselamatan. Tujuan penelitian adalah menentukan nilai subkritikalitas dalam keadaan normal atau kondisi kecelakaan di gedung reaktor dan di gudang penyimpan bahan bakar bekas. Perhitungan dilakukan dengan paket program MCNP6.1. Kejadian kecelakaan yang dipilih adalah masuknya air ke dalam cask, cask terendam air dan kombinasi cask terendam air dan kejadian gempa. Hasil perhitungan menunjukkan bahwa nilai maksimum keff (3σ) untuk cask di gedung reaktor dan di gudang penyimpan bahan bakar bekas masing-masing adalah 0,47510 dan 0,19214. Nilai ini masih jauh dari batas 0,95. Hasil perhitungan menunjukkan bahwa cask penyimpan bahan bakar bekas tetap dalam keadaan selamat meski terjadi kombinasi 2 kejadian eksternal.Kata kunci: elemen bahan bakar bekas tipe pebble, HTGR, subkritikalitas, MCNP6.1, RDE
A key issue contributing to the success of NPP technology is the safe handling of radioactive waste, particularly spent nuclear fuel. According to the IAEA safety standard, the spent fuel must be stored in interim wet storage for several years so the radiation and the decay heat of the spent fuel will decrease to the safe limit values, after which the spent fuel can be moved to dry storage. In this study, we performed a theoretical analysis of heat removal by natural convection airflow in spent nuclear fuel dry storage. The temperature difference between the air inside and outside dry storage produces an air density difference. The air density difference causes a pressure difference, which then generates natural airflow. The result of the theoretical analysis was validated with simulation software and experimental investigation using a reduced-scale dry storage prototype. The dry storage prototype consisted of a dry cask body and two canisters stacked to store materials testing reactor (MTR) spent fuel, which generates decay heat. The cask body had four air inlet vents on the bottom and four air outlet vents at the top. To simulate the decay heat from the spent fuel in the two canisters, the canisters were wrapped with an electric wire heater that was connected to a voltage regulator to adjust the heat power. The theoretical analysis results of this study are relatively consistent with the experimental results, with the mean relative deviation (MRD) values for the prediction of air velocity, the heat rate using natural airflow, and the heat rate using the thermal resistance network equation are +0.76, −23.69, and −29.54%, respectively.
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