INVESTIGATION OF AG AND PD FISSION PRODUCTS PENETRATION DEPTH IN ZRC LAYER OF HTGR TRISO FUEL PARTICLE USING SRIM/TRIM MONTE CARLO SIMULATION

High-temperature gas-cooled Reactors (HTGRs) is one type of Generation IV reactor that uses TRISO (tri-structural isotropic) coated-fuel particles (CFP) for containment of radioactive fission products, which is produced from the fission reaction of UO2 fuel. ZrC has been proposed to be the main barrier for containing fission products either as a replacement of the SiC layer or as an additional layer of the TRISO fuel particle to overcome the corrosion issue of SiC because of interaction with the fission product of silver (Ag) and palladium (Pd). ZrC is an excellent material because it has good physical and nuclear properties, i.e., high corrosion-resistant, excellent thermal shock resistance and a small cross-section for neutron capture. ZrC is expected to provide a better barrier against Ag and Pd diffusion attacks than SiC. However, ZrC is very challenging to manufacture, so it depends on factors such as microstructure, chemical composition and interactions, morphology and impurities. Many attempts have been made to study the interaction phenomena of Ag and Pd with ZrC that cause corrosion. Here, the penetration depth of those two fission products was studied using SRIM (Stopping and Range of Ions in Matter) /TRIM (TRansport of Ions in Matter) for simulation with 0.1-10 MeV of kinetic energies. The results provide detailed information about the Ag/ZrC and Pd/ZrC Ion Ranges and Doses. In addition, Ag and Pd’s products of the depth and concentration within ZrC were observed as important first steps in understanding the corrosion phenomena of her ZrC layers in TRISO particles

TEMPERATURE MAPPING OF TRIGA 2000 REACTORS AT 500 kW POWER WITH 105 CONFIGURATIONS OF PRE-RESHUFFLING AND POST-RESUFFLING FUEL

The continuity of operation of the TRIGA 2000 reactor is thermohydraulically determined by the temperature of the fuel and the temperature of the primary cooling water in the reactor core. Currently, the operation of the TRIGA 2000 reactor at a power of 500 kW using a fuel 105 configuration in the core has caused boiling and the formation of bubbles in the reactor core, thereby reducing neutron moderation. The core of reactor is unable to achieve more power. One of the efforts that can be made to reduce the core temperature of the TRIGA 2000 reactor is by reshuffling the fuel inside the core, including using 105 fuels but shifting the position of the fuel to a different ring. In this research, thermohydraulic analysis has been carried out using the FLUENT program package for the configuration of 105 fuels in the pre-reshuffling and post-reshuffling core. Based on the results of the study, it is known that at 500 kW of power with a pre-reshuffling fuel configuration, there are positions for sub-cooled boiling in B3, B4, B5, B6, D4, D6, D8, D10, and D12. The hottest channel is at B5 with a maximum fuel cladding surface temperature of 126.41 °C, and a cooling water saturation temperature in the core of 112.4 °C. Meanwhile, for the post-reshuffling fuel configuration there is no position for sub-cold boiling in the core, where the maximum temperature is at B2 with the maximum fuel cladding surface temperature of 93.55 °C. Thus, it is concluded that the TRIGA 2000 reactor with a configurations of 105 post-reshuffling fuel can be operated at of 500 kW without boiling or at least not occurring sub-cold boiling

STUDY OF PLUTONIUM UTILIZATION IN AP1000 REACTOR USE SRAC2006 AND JENDL 3.3 NUCLEAR DATA LIBRARY

Preliminary Study of Plutonium Utilization in AP1000 Reactor Use SRAC 2006 and JENDL 3.3 has been conducted. Nuclear energy, especially for nuclear reactor, become important this day because the need of energy will increase along with the increasing of human population, the advanced technology and economic. The more nuclear reactor operated the more existence of plutonium stockpile. This study evaluated the standard of Westinghouse AP1000 reactor and ZrB2 as Integral Fuel Burnable Absorber (IFBA). Different fuel compositions of assembly type were analyze in by using SRAC 2006 code system with JENDL 3.3 nuclear data library. This study aiming to compare the neutronics characteristics of an UO2 and an (U,Pu)O2 assembly designs. Some results of the study show that optimal criticality of the fuel assembly can be accomplished by using 5% enrichment of U-235 for UO2 fuel and 9% plutonium fraction for (U,Pu)O2 fuel assembly

ANALYSIS OF EYE LENS EFFECTIVE DOSE OF WORKERS DURING CARDIAC CATHERIZATION EXAMINATION IN A CATHLAB ROOM

In the current era, many radiological imaging technologies are used, one of which is the C-Arm equipment technology, which is capable of imaging three-dimensional patient objects in real time using fluoroscopy techniques. This study aims to obtain and evaluate the effective radiation dose received around the eye organs of radiation workers during cardiac catheterization. The main tool uses the C-Arm modality, and the TLD detector chip. The object of observation is the area around the eye organs of radiation workers consisting of one doctor and one nurse. Catheterization was carried out on 10 patients with different examination times. Research results: in the area around the eyes without protective glasses; the effective dose received by doctors' eyes ranges from 0.0011 to 0.0054 mSv, while for nurses it ranges from 0.0010 mSv to 0.0025 mSv. It was found that the effective dose received around the eye organs was lower than the dose value set by Perka Bapeten No. 8 of 2011 which is 20 mSv and ICRP 203 which is 500 mSv per year

ANALYSIS ON HEAT LOSS IN WATER HEATING TANK BASED ON TEMPERATURE SETTING VARIATION DURING NATURAL CIRCULATION FLOW USING FASSIP-02 TEST LOOP

Based on the nuclear reactor accident in Fukushima due to an earthquake that caused a tsunami that turned off the electrical system for the cooling water pump. So that the reactor core melts due to the failure of the active cooling system, based on this incident, a passive cooling system that does not require external energy for emergency cooling is needed to prevent such an accident. Experimental research on the passive cooling system based on natural circulation flow to improve reactor thermal management performance during an accident has been done using the FASSIP-02 Test. One of the important components of the simulation of heat sources in nuclear power plants is the water heating tank (WHT). This study aimed to obtain the analysis results of heat absorption in water in WHT and the value of heat loss. The research method was carried out by calculating changes in the internal energy of water in WHT and heat loss for variations in water temperature settings from temperatures of 40 oC, 50 oC, and 60 oC in WHT at steady-state conditions for 5 hours experiment. The results showed that the entire surface of the tank resulted in heat losses of 10.85 kW, 9.2 kW, and 8.37 kW, which occurred at temperature settings of 40 oC, 50 oC, and 60 oC as a whole

THERMAL POWER CALIBRATION OF TRIGA 2000 RESEARCH REACTOR

The thermal power calibration of the TRIGA 2000 research reactor is very important to get the accuracy of power and neutron flux. The TRIGA 2000 research reactor in Bandung has undergone fuel reshuffling and therefore requires thermal power calibration. Thermal power calibration has been conducted by calorimetric method; it is performed at 100 kW – 500 kW using 7 thermocouples connected to a data logger. The computed average power was lower than the indicated power shown in the control room for all power generation. When observing channel 1 for each power generation, a higher precision can be seen at 500 kW since the calibration process was carried out sequentially from 100 kW to 500 kW while the stirring process was continually operated. The treatment led to a uniform temperature distribution over time. Each measurement channel exhibited inconsistent deviations, indicating that certain power levels had better accuracy in some channels compared to others. This demonstrates that the accuracy of power calculations is not determined by the measurement position

RANCANG-BANGUN SISTEM PENCUPLIK UDARA MODEL PORTABLE-ADJUSTABLE

Dalam upaya melindungi pekerja, masyarakat, serta lingkungan, perlu dilakukan monitoring dan pengawasan secara berkala mencakup seluruh aspek keselamatan di lingkungan fasilitas nuklir yang mengacu kepada peraturan perundang-undangan. Monitoring perlu dilakukan untuk mengetahui nilai radioaktivitas komponen lingkungan yang berpotensi berdampak bagi kualitas lingkungan sehingga dijadikan dasar evaluasi pengelolaan lingkungan. Udara merupakan salah satu komponen lingkungan yang dipantau dengan pengukuran secara tidak langsung di laboratorium. Monitoring terhadap komponen udara memerlukan sistem sarana dan peralatan pendukung lainnya. Selain itu, sarana yang ada dinilai kurang praktis mengingat kondisi dan kendala di lapangan. Oleh karena itu, pada tulisan ini dijelaskan mengenai pemenuhan sarana yang diperlukan dengan cara merancang sistem pencuplik udara yang cukup ekonomis, desain model portable-adjustable sehingga selain mendapatkan sarana yang relatif murah, juga memberikan kemudahan untuk dapat meminimalisasi beberapa kendala yang dihadapi di lapangan. Perancangan dilakukan dengan cara mengidentifikasi kendala yang dihadapi di lapangan, menentukan bagian-bagian sistem yang diperlukan, lalu membangun sistem dengan memanfaatkan komponen yang telah tersedia dan mudah didapat di pasaran. Telah dirancang dan dibuat sistem alat pencuplik udara model portable-adjustable. Sistem alat telah dioperasikan dan digunakan untuk kegiatan monitoring komponen udara

KAJIAN PENGARUH TEGANGAN LISTRIK TERHADAP KERUSAKAN KOPLING POMPA PENDINGIN SEKUNDER SELATAN REAKTOR TRIGA 2000 BANDUNG

Telah dilakukankajian pengaruh tegangan listrik terhadap kerusakan kopling pompa pendingin sekunder selatan, kajian ini dilakukan tujuannya adalah untuk mengetahui apakah tegangan listrik  berpengaruh terhadap  kerusakan kopling pompa, sehingga harus diganti dengan yang baru. Kegiatan penggantiankopling pompa  dilakukan diakhir tahun 2017. Metoda kajian menggunakan Unbalance Voltage, kegiatan dimulai dengan mengukur besaran tegangan listrikdi tiga kabel yaitu(V(R-S), V(R-T), V(S-T))yang telah melewati Molded Case Circuit Breaker (MCCB) dan dua kontaktor, dilakukan selama 2 jam.Untuk mengetahui keseimbangan dari tegangan listrik yang terukur, maka data hasil pengukuran tegangan yang terkumpul, kemudian dihitung menggunakan rumus NEMA Unbalance Voltage. NEMA memberikan rekomendasi: motor dapat dioperasikan secara normal pada kapasitas rated jika unbalance voltage ≤ 1%. Hasil perhitungan nilai persentase unbalance voltage ada yang mencapai nilai sebesar: 2,17%. Hasil tersebut menunjukan bahwa nilai Unbalance voltage pada  pompa pendingin sekunder selatan tidak berpengaruh  terhadap kerusakan kopling pompa tetapi berpengaruh kepada kenaikan suhu motor pompa sehingga menyebabkan  panas yang berlebih (overheatin) Dapat disimpulkan bahwa nilai Unbalance voltage pada panel catu daya listrik system pendingin sekunder bukan sebagai penyebab kerusakan kopling pompa, namun kerusakan kopling disebabkan karena adanya penurunan kinerja akibat umur pakai komponen

KAJIAN PENETAPAN NILAI PEMBATAS DOSIS PEKERJA RADIASI DI PSTNT

Telah dilakukan Kaji Ulang Penetapan Nilai Pembatas Dosis (NPD) Pekerja Radiasi yang telah berlaku sejak tahun 2014 di PSTNT. Beroperasinya kembali reaktor TRIGA 2000 secara rutin sejak tahun 2017 meningkatkan volume kegiatan laboratorium RISB dan AAN. Peningkatan volume kegiatan tersebut menyebabkan peningkatan terimaan paparan radiasi dan berakibat pada meningkatnya terimaan dosis akumulasi Pekerja Radiasi di PSTNT, sehingga perlu dilakukan kaji ulang terhadap NPD Pekerja Radiasi di PSTNT. Tujuan dari kaji ulang NPD adalah untuk memperbaharui nilai pembatas dosis yang telah ada sebelumnya (15 mSv/tahun). Kaji ulang dilakukan dengan menginventarisasi: jenis kegiatan dengan sumber bahaya potensial di reaktor dan laboratorium radiasi, terimaan dosis pekerja radiasi selama 4 tahun (2015 – 2018) dan membandingkan jenis kegiatan dengan sumber bahaya potensial dengan terimaan dosis pekerja radiasi rata-rata. Peningkatan kegiatan di PSTNT mengakibatkan semakin meningkatnya terimaan dosis akumulasi rata- rata, dosis maksimum tahunan dan dosis kolektif dari tiap bidang di PSTNT. Meningkatnya dosis tersebut menunjukkan kecenderungan Nilai Batas dosis (NBD) dapat dilampaui dan resiko radiasi bagi para pekerja juga akan semakin meningkat. Untuk menurunkan resiko radiasi bagi Pekerja Radiasi maka nilai NBD tidak boleh dilampaui sehingga perlu dilakukan penurunan NPD dari 15 mSv/tahun menjadi 10 mSv/tahun

THE ANALYSIS OF LAPSE RATE PROFILE IN THE SITE CANDIDATE OF NUCLEAR POWER PLANT (NPP) AT GOSONG BEACH, BENGKAYANG REGENCY– WEST KALIMANTAN

The lapse rate profile in the site candidate for the Nuclear Power Plant (NPP) at Gosong Beach Bengkayang, has been investigated to obtain a description of the lability of the atmosphere and upper air as part of a meteorological aspect safety study in the plan to develop a NPP site. The study of the lapse rate was carried out using air data on the reanalysis of the Global Data Assimilation System (GDAS) by extracting air temperature data at each altitude level so as to obtain a lapse rate of up to 25 km. Daily data was processed during 2021 and transformed in the monthly average profile data to describe the lapse rate profile in January – December 2021. Tropopause was identified with average altitude about 16.6 km and stratosphere at 20.5 km with a lapse rate about -0.21 ℃/100 m. The surface layer to 200 m have lapse rate from 0.7 ℃/100 m - 0.9 ℃/100 m at 00.00 Universal Time Coordinated (UTC) and 0.5 ℃/100 m -0.6 ℃/100 m at 12.00 UT