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"This study proposes a method of optimizing the dry storage design for nuclear-spent fuel from the G.A. Siwabessy research reactor at National Nuclear Energy Agency of Indonesia (BATAN). After several years in a spent fuel pool storage (wet storage), nuclear spent fuel is often moved to dry storage. Some advantages of dry storage compared with wet storage are that there is no generation of liquid waste, no need for a complex and expensive purification system, less corrosion concerns and that dry storage is easier to transport if in the future the storage needs to be sent to the another repository or to the final disposal. In both wet and dry storage, the decay heat of spent fuel must be cooled to a safe temperature to prevent cracking of the spent fuel cladding from where hazardous radioactive nuclides could be released and harm humans and the environment. Three optimization scenarios including the thermal safety single-objective, the economic single-objective and the multi-objective optimizations are obtained. The optimum values of temperature and cost for three optimization scenarios are 317.8K (44.7°C) and 11638.1 US$ for the optimized single-objective thermal safety method, 337.1K (64.0°C) and 6345.2 US$ for the optimized single-objective cost method and 325.1K (52.0°C) and 8037.4 US$ for the optimized multi-objective method, respectively."

"Penelitian ini bertujuan untuk memberikan solusi penyimpanan Bahan Bakar Nuklir Bekas (SNF) di Indonesia. Karena keterbatasan ruang pada penyimpanan tipe basah, maka penelitian ini merancang, mensimulasikan, melakukan eksperimen, dan menghitung biaya pembuatan penyimpanan dry cask storage secara simultan. Dalam studi ini, desain dry cask storage dioptimalkan dengan menggunakan dua objective functions secara bersamaan yaitu fungsi keselamatan (yang menggabungkan parameter kekritisan, proteksi radiasi dan penghilangan panas) dan fungsi biaya. Perhitungan optimasi kemudian divalidasi dan dianalisis dengan data eksperimen dari prototipe dry cask storage. Dengan menentukan decision variables dan constraints, kemudian memasukkannya ke dalam Matlab software, diperoleh tiga pilihan hasil optimasi, safety optimized, cost optimized dan multi-objective optimized. Dalam multi-objective optimized, desain penyimpanan kering yang optimal diperoleh untuk radius luar beton dan timbal (Pb) masing-masing sebesar 0,06 m dan 0,51 m. tinggi ventilasi dan lebar masing-masing sebesar 0,15 m dan 0,5 m, dan perbedaan ketinggian ventilasi sebesar 2,43 m. Untuk kelima variabel diatas, nilai optimum temperatur permukaan kanister adalah 66,8 °C dan biaya yang dibutuhkan untuk pembuatan dry cask storage adalah $147,827. Ketebalan material yang dibutuhkan didapatkan nilai yang paling optimum untuk Pb 0,06 m dan beton 0,51 m, dari validasi menggunakan MicroShield software didapatkan paparan permukaan penyimpanan dry cask sebesar 104,8 mR/jam sehingga masih dalam batas aman dari nilai maksimum yang ditentukan yaitu 160 mR/jam. Demikian pula dari simulasi suhu permukaan tabung menggunakan Ansys Fluent software untuk kelima variabel di atas, nilai suhu permukaan mendekati perhitungan yang persamaannya dimasukkan ke dalam Matlab software. Validasi menggunakan data eksperimen dari prototipe dry cask storage dan juga perhitungan manual diperoleh nilai temperatur yang juga relatif mendekati hasil optimasi, yaitu 45,2 °C untuk temperatur dari eksperimen dan 50.2 °C untuk temperatur perhitungan teori. Hasil nilai optimasi terpilih dengan tetap menjaga keamanan termal menunjukkan bahwa acuan dalam pembuatan desain dengan skala 1:1 dapat menggunakan estimasi untuk keamanan termal T1, untuk jenis SNF Materials Testing Reactor (MTR) dengan umur setelah sepuluh tahun disimpan pada jenis penyimpanan tipe basah.

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This research aims to provide a solution for Indonesia's spent nuclear fuel (SNF) storage. Due to the limited storage space in wet type storage, this research designs, simulates, conducts experiments, and calculates the cost of making dry cask storage simultaneously. In this study, the dry storage design was optimized by using two objective functions simultaneously: safety function (which combines criticality parameters, radiation protection and heat removal) and cost function. The optimization calculations were then validated and analyzed with experimental data from the dry cask storage prototype. By determining the decision variables and constraints and then inputting them into the Matlab software, three choices of optimization results are obtained, safety optimized, cost optimized and multi-objective optimized. In multi-objective optimized, the optimum dry storage design is obtained for the concrete outer- and lead (Pb) outer-radius of 0.06 m and 0.51 m respectively. vent-height, -widt by 0.15 m and 0.5 m respectively, and a vent elevation difference of 2.43 m. For the five variables above, the optimum value for the canister surface temperature is 66.8 °C and the cost required to make dry storage is $ 147,827.The required thickness of the material obtained the most optimum value for Pb 0.06 m and concrete 0.51 m, from validation using MicroShield software, it was obtained that the dry cask storage surface exposure was 104.8 mR/h so that it is still within the safe limit of the maximum value specified, which is 160 mR/h. Similarly, from the simulation of canister surface temperature using Ansys Fluent for the five variables above, the surface temperature value is close to the calculation whose equations are inputted into matlab. Validation using experimental data from the dry storage prototype and also manual calculations obtained temperature values which are also relatively close to the optimization results, 45.2 °C for temperature from experiments and 50.2 °C for temperature from manual calculations. The result of the selected optimization value while maintaining thermal safety indicates that reference in making designs with a scale of 1:1 can use the estimate for the thermal safety of T1, for the type of SNF Materials Testing Reactor with age after ten years stored in the wet storage type."

"Penelitian ini membahas optimasi sistem ventilasi dan tata udara untuk Instalasi Penyimpanan Sementara Bahan Bakar Nuklir Bekas dari Reaktor Riset G.A. Siwabessy. Keterbaruan dari penelitian ini adalah didapatkannya beberapa persamaan yang diperlukan. Optimasi yang dilakukan pada penelitian ini selain untuk Instalasi Penyimpanan Sementara Bahan Bakar Nuklir Bekas yang sudah ada juga untuk Instalasi Penyimpanan Sementara Bahan Bakar Nuklir Bekas tipe kering yang saat ini belum dibangun. Hasil dari penelitian ini menunjukkan bahwa beberapa persamaan yang dikembangkan pada penelitian ini bisa dibuktikan mendekati hasil eksperimen.

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This study discusses an optimization of ventilation and air conditioning systems for the Interim Storage of Spent Nuclear Fuel Fuel from the G.A Siwabessy Research Reactor. The novelty of this research is to gain several equations needed for the optimization. The optimization in this study is in addition to the existing wet interim storage also for a dry storage that is currently still not built. The results of this study indicate that the equations developed in this study are consistent with the experimental results."

"ABSTRAKStruktur pengungkung reaktor nuklir yang berfungsi sebagai proteksi terhadap aksi lingkungan serta penahan gaya gempa merupakan struktur utama dengan faktor keutamaan tinggi. Penelitian ini bertujuan untuk memastikan kemampuan serta daya tahan struktur pengungkung reaktor nuklir terhadap gempa serta memperkirakan dampak penyebaran kontaminasi radiasi ke lingkungan pasca-kecelakaan. Gempa Pariaman, Padang digunakan sebagai input gempa dengan diskalakan pada gempa rencana periode ulang 10.000 tahun.Hasil penelitian menunjukkan struktur reaktor merupakan struktur sangat kaku dengan drift sangat kecil. Kondisi inelastis struktur tertutup baru terjadi ketika ditingkatkan hingga 2 kali. Struktur tertutup serta struktur terbuka yang diperkuat dengan kolom mampu bertahan hingga 1,5 kali gempa rencana sebelum melampaui tegangan ijinnya. Alur paparan penyebaran kontaminasi dapat melalui tanah (Groundshine), atmosfer (Cloudshine), hirupan (inhalation), kontaminasi makanan (ingestion) hingga aliran air (Aquatic Pathway).

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ABSTRACT

Nuclear reactor containment structure that serves as protection against environmental action and retaining the force of the earthquake is the main structure that highly safety factor. This study aims to ensure the capability and durability of a nuclear reactor containment structure against earthquakes and estimates the impact of the spread of radiation contamination into the environment. Earthquake Pariaman, Padang is used as input to the scaled earthquake with return period of 10,000 years.

The results showed the reactor structure is very rigid structure with very little drift. The inelastic behavior of enclosed structure occurs when earthquake increased by 2 times. Structures closed or open structures with columns can last up to 1.5 times before the earthquake exceeded the stress limit. Chronology of exposure spread of contamination can be Groundshine, atmosphere (Cloudshine), inhalation, contamination of food (ingestion), or to the flow of water (Aquatic Pathway)."

"Indonesia is expected to apply the strategy of long term storage for spent fuel generated from the Serpongresearch reactor. The capacity of existing interim storage for spent fuel (ISSF) facility in principle able toaccommodate all the spent fuel generated Serpong research reactor, but it must consider long-term conditions ofthe spent fuels and the wet storage facilities. This long-term strategy requires special attention to someparameters dealing with the water chemsitry and the degradation of the materials. Besides it is necessary to builtreserve space to deal with emergencies. After the Serpong reactor decommissioned, it is recommended to buildthe new dry storage to accomodate all of the spent fuel in another location since the Serpong area will be verydense residential in the decades to come. The most realistic future back-end scenario is if Indonesia has nuclearpower plants (NPP), then the disposal of the spent fuel generated from research reactor in the future can be donein one location with commercial spent fuel from the NPP."

"The nuclear reactor is a source of primary radiation. Safety in the operation of thereactor protection system is implemented by the scram action. However scramwhich often happened is the failure of the operation and poses other risks. Thepurpose of this study is to determine the trigger and the reasons for scram withanalytic descriptive method through interviews, observation and documentreview, combined with the preparation of fault tree analysis. The obtained resultsshow that trigger of scram are, the neutron flux density is too high, too fastoperation period, unbalanced load, and the occurrence of positive transientreactivity. While the basic causes of the scram incident are, too fast control rod,irradiation sample composition, corrosion, weakening of the power supplydetector, emptying of the beam tube, and unbalanced neutron flux. Then scram asanticipation of the accident is also the feedback of operating experience forupdating on safety assessment so that failure of the operation can be minimized.Keywords The safe operation of nuclear reactor, reactor protection system, cause of theincident, suddenly shutdown of reactor.

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Reaktor nuklir merupakan sumber radiasi primer. Keselamatan dalampengoperasian reaktor diterapkan oleh sistem proteksi dengan tindakan scram.Namun scram yang sering terjadi merupakan kegagalan operasi dan menimbulkanrisiko lain. Tujuan dari penelitian ini adalah untuk mengetahui pemicu danpenyebab dasar terjadinya scram dengan metode deskriptif analitik melaluiwawancara, observasi, dan telaah dokumen yang dipadukan dengan penyusunanfault tree analysis. Maka diperoleh hasil bahwa pemicu timbulnya scram, yaitukerapatan fluks neutron terlalu tinggi, periode pengoperasian terlalu cepat,pembebanan atau daya di teras tidak merata, dan terjadinya transien reaktivitaspositif. Sedangkan penyebab dasar kejadian scram, yaitu kenaikan batang kendaliterlalu cepat, kesalahan komposisi sampel iradiasi, korosi, pelemahan catu dayadetektor, pengosongan tabung berkas, dan fluks neutron tidak merata. Maka scramsebagai antisipasi kecelakaan juga menjadi umpan balik pengalaman operasiuntuk rekomendasi pemutakhiran penilaian keselamatan sehingga kegagalanoperasi dapat diminimalisir.Kata kunci:Keselamatan operasi reaktor nuklir, Sistem proteksi reaktor, Penyebab kejadian,Scram reaktor."

"Radionuklida [¹⁷⁷Lu]Lu bebas pengemban adalah sediaan radionuklida [¹⁷⁷Lu]Lu dengan aktifitas spesifik yang sangat tinggi namun dibutuhkan pemisahan yang sangat sulit untuk memperolehnya. Pada penelitian ini dikembangkan metode produksi [¹⁷⁷Lu]Lu bebas pengemban dari aktivasi tidak langsung isotop Ytterbium alam menggunakan metode pemisahan kromatografi penukar ion dengan fasa diam resin Dowex W50 X8 dan campuran eluen alpha Hydroxyisobutiric Acid (α-HIBA) dan HCl. Hasil penelitian menunjukan bahwa reaktor nuklir G.A Siwabessy telah mampu menghasilkan [¹⁷⁷Lu]Lu sebanyak 296 MBq/10 mg sampel Yb₂O₃. Dua metode pemisahan spesifik diperoleh, pertama dengan menahan radionuklida [¹⁷⁷Lu]Lu di dalam resin sedangkan ion dan radionuklida pengotor keluar dari kolom menggunakan campuran eluen HCl 0,25 M dan ɑ-HIBA 0,1 M yang dapat mengeluarkan pengotor [¹⁷⁵Yb]Yb  sekitar 1,6 x 10^-3 % yield/ml dan [¹⁶⁹Yb]Yb sekitar 4,2 % yield/ml. Metode kedua didapatkan dengan menggunakan peningkatan konsentrasi eluen α-HIBA 0,15 M yang menyebabkan [¹⁷⁷Lu]Lu keluar dari kolom sedangkan pengotor tetap berada di dalam kolom dengan kemurnian [¹⁷⁷Lu]Lu sekitar 81,9 % dan aktifitas spesifik (1,163 GBq/mg). Faktor peningkatan konsentrasi HCl diatas 0,25 M pada eluen menyebabkan penurunan selektifitas pemisahan [¹⁷⁷Lu]Lu dari matriks ytterbium. Sedangkan faktor peningkatan temperatur elusi 50 ^oC dapat menaikan selektifitas pemisahan dengan menahan lebih baik [¹⁷⁷Lu]Lu di dalam resin.

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No-carrier added [¹⁷⁷Lu]Lu radionuclide is a [¹⁷⁷Lu]Lu radionuclide preparation with very high specific activity but requires very difficult separation to obtain it. In this research, a carrier-free [¹⁷⁷Lu]Lu production method was developed from indirect activation of natural Ytterbium isotopes using an ion exchange chromatography separation method with a Dowex W50 X8 resin and mixed eluent alpha hydroxyisobutyric acid (α-HIBA) and hydrochloric acid (HCl). The research results showed that the G.A Siwabessy nuclear reactor was able to produce [¹⁷⁷Lu]Lu as much as 296 MBq/10 mg Yb₂O₃ sample. Two specific separation methods were obtained, first by retaining the [¹⁷⁷Lu]Lu radionuclide in the resin while the impurity ions and radionuclides come out of the column using a mixture of 0.25 M HCl and 0.1 M ɑ-HIBA eluents which can remove [¹⁷⁵Yb]Yb impurities around 1.6 x 10-3 % yield/ml and [¹⁶⁹Yb]Yb around 4.2 % yield/ml. The second method was obtained by using an increase in the eluent concentration of 0.15 M α-HIBA which caused [¹⁷⁷Lu]Lu to come out of the column while the impurities remained in the column with a [¹⁷⁷Lu]Lu purity of around 81.9% and specific activity (1.163 GBq/mg ). The increasing factor of HCl concentration above 0.25 M in the eluent causes a decrease in the selectivity of [¹⁷⁷Lu]Lu separation from the ytterbium matrix. Meanwhile, increasing the elution temperature by 50 ^oC can increase separation selectivity by better retaining [¹⁷⁷Lu]Lu in the resin."

"[ABSTRAKLimbah Bahan Bakar Nuklir Bekas (BBNB) merupakan salah satu limbah yang dihasilkan dalam pengoperasian reaktor nuklir. Limbah ini masih menghasilkan produk fisi dan panas hasil reaksi yang masih tinggi sehingga perlu dikelola dengan baik agar efek radiasi yang ditimbulkan tidak keluar di lingkungan. Penelitian ini akan dilakukan pemodelan panas peluruhan pada penyimpanan kering BBNB bentuk pebble dengan tipe storage tank yang telah digunakan pada reaktor HTR 10 menggunakan software ORIGEN-ARP. Dengan computational fluid dynamics (CFD) menggunakan Comsol Multiphysics maka pengaruh kecepatan udara pendingin dan ketebalan lapisan pengungkung terhadap profil suhu di setiap segmen storage dapat diketahui sehingga keselamatan penyimpanan BBNB pada aspek suhu dapat dianalisis. Dari hasil perhitungan dapat diketahui bahwa panas peluruhan yang dihasilkan oleh BBNB setelah keluar dari reaktor sebesar 620,2260 watt. Panas peluruhan tersebut semakin menurun seiring dengan lamanya waktu penyimpanan. Ketebalan beton tidak terlalu berpengaruh terhadap penurunan suhu di storage tank. Hal ini disebabkan oleh konduktivitas panas beton yang rendah sehingga laju perpindahan panas di setiap variasi ketebalan tidak berbeda secara signifikan. Ketebalan stainless steel berpengaruh terhadap gradien perubahan suhu pada storage tank. Semakin tipis stainless steel maka semakin banyak laju panas yang dialirkan dari grafit ke beton, sehingga suhu pada beton semakin besar. Semua hasil simulasi pada berbagai kondisi memenuhi syarat parameter suhu maksimum keselamatan.

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ABSTRACT

Nuclear Fuel Waste is one of waste generated in operation of nuclear reactors. This waste is still producing fission products and heat of reaction that need to be managed properly so the effects of radiation emitted do not expose to environment. This research will be carried out modeling the decay heat in dry storage of pebble nuclear spent fuel with the type of storage tanks that have been used in the reactor HTR 10 using ORIGEN-ARP software. The effects of cooling air velocity and confinement layer thickness on temperature profile in every segment of storage can be determined with computational fluid dynamics (CFD) using Comsol Multiphysics so the safety of nuclear spent fuel storage on temperature aspects can be analyzed. Based on the calculation results can be seen that the decay heat generated by nuclear spent fuel after coming out from the reactor is 620.2260 watts. The decay heat decreases as the length of storage time. Concrete thickness does not significantly affect the declining temperature gradient in the storage tank. This is caused by the low thermal conductivity of concrete so the heat transfer rate in each variation of thickness is not different significantly. Stainless steel thickness affects the declining temperature gradient. Thinner the thickness of the stainless steel is used, greater the reduction of temperature gradient so equilibrium temperature of storage tank can be quickly achieved. All simulation results under various conditions compliy with the maximum temperature parameters of safety.;Nuclear Fuel Waste is one of waste generated in operation of nuclear reactors. This waste is still producing fission products and heat of reaction that need to be managed properly so the effects of radiation emitted do not expose to environment. This research will be carried out modeling the decay heat in dry storage of pebble nuclear spent fuel with the type of storage tanks that have been used in the reactor HTR 10 using ORIGEN-ARP software. The effects of cooling air velocity and confinement layer thickness on temperature profile in every segment of storage can be determined with computational fluid dynamics (CFD) using Comsol Multiphysics so the safety of nuclear spent fuel storage on temperature aspects can be analyzed. Based on the calculation results can be seen that the decay heat generated by nuclear spent fuel after coming out from the reactor is 620.2260 watts. The decay heat decreases as the length of storage time. Concrete thickness does not significantly affect the declining temperature gradient in the storage tank. This is caused by the low thermal conductivity of concrete so the heat transfer rate in each variation of thickness is not different significantly. Stainless steel thickness affects the declining temperature gradient. Thinner the thickness of the stainless steel is used, greater the reduction of temperature gradient so equilibrium temperature of storage tank can be quickly achieved. All simulation results under various conditions compliy with the maximum temperature parameters of safety., Nuclear Fuel Waste is one of waste generated in operation of nuclear reactors. This waste is still producing fission products and heat of reaction that need to be managed properly so the effects of radiation emitted do not expose to environment. This research will be carried out modeling the decay heat in dry storage of pebble nuclear spent fuel with the type of storage tanks that have been used in the reactor HTR 10 using ORIGEN-ARP software. The effects of cooling air velocity and confinement layer thickness on temperature profile in every segment of storage can be determined with computational fluid dynamics (CFD) using Comsol Multiphysics so the safety of nuclear spent fuel storage on temperature aspects can be analyzed. Based on the calculation results can be seen that the decay heat generated by nuclear spent fuel after coming out from the reactor is 620.2260 watts. The decay heat decreases as the length of storage time. Concrete thickness does not significantly affect the declining temperature gradient in the storage tank. This is caused by the low thermal conductivity of concrete so the heat transfer rate in each variation of thickness is not different significantly. Stainless steel thickness affects the declining temperature gradient. Thinner the thickness of the stainless steel is used, greater the reduction of temperature gradient so equilibrium temperature of storage tank can be quickly achieved. All simulation results under various conditions compliy with the maximum temperature parameters of safety.]"