PRELIMINARY DEVELOPMENT OF RADIONUCLIDES RELEASE OF INDIVIDUAL DOSE CODE PROGRAM FOR RADIATION MONITORING PURPOSES

Environmental radiation monitoring is one of the important efforts in protecting society and the environment from radiation hazards, both natural and artificial. The presence of three nuclear research reactors and plans to build a nuclear power plant reactor prompted Indonesia to prepare a radiation monitoring system for safety and security (SPRKK). The goal of the study is to provide an appropriate method for developing radiation monitoring system to support the development of nuclear power plant in the near future. For this preliminary study, the author developed a code program using Gaussian distribution model approach for predicting radionuclide release and individual dose acceptancy by human being within 16 wind directions sectors and up to 50 km distance. The model includes estimation of source term from the nuclear installation, release of radionuclides source into air following Gaussian diffusion model, some of the release deposit to the land and entering human being through inhalation, direct external exposure, and resuspension, and predicted its accepted individual dose. This model has been widely used in various code program such as SimPact and PC-Cosyma. For this study, the model will be validated using SimPact code program. The model has been successfully developed with less than 5% deviation. Further study will be done by evaluating the model with real measuring data from research reactor installation and prepare for interfacing with real time radiation data acquisition and monitoring as part of radiation monitoring system during normal and accident condition.

PARTICLE SWARM OPTIMIZATION BASED PROBABILISTIC NEURAL NETWORK FOR CLASSIFICATION OF SEVERE ACCIDENT OF NUCLEAR REACTOR

Due to its danger and complexity, the identification and prediction of major severe accident scenarios from an initiating event of a nuclear power plant remains a challenging task. This paper aims to classify severe accident at the Advanced Power Reactor (APR) 1400, which includes the loss of coolant accidents (LOCA), total loss of feedwater (TLOFW), station blackout (SBO), and steam generator tube rupture (SGTR) using a standard probabilistic neural network (PNN) and Particle Swarm Optimization Based Probabilistic Neural Network (PSO PNN). The algorithm has been implemented in MATLAB. The experiment results showed that supervised PNN PSO could classify severe accident of nuclear power plant better than the standar PNN.

MAP OF RADIOISOTOPE PRODUCTION AND BATAN RESEARCH REACTOR UTILIZATION

Currently, Indonesia through BATAN is operating three research reactors, namely the RSG-GAS reactor with the power of 30 MWt at Puspiptek south Tangerang (the first criticality in 1987), the TRIGA 2000 reactor with the power of 2 MW in Bandung which the first criticality in 1965 with the power of 250 kW, was increased to 1 MW in 1971, and further upgraded to 2 MW in 2000. Beside that, there is Kartini reactor with a power of 100 kW located in Yogyakarta (first criticality in 1979). These reactors are quite old, and in accordance with Bapeten regulations, have carried out the first periodic safety review, to obtain a reactor license for the next 10 years of operation. In line with this, one of BATAN's current national research programs is to increase the production of radioisotopes and radiopharmaceuticals, where reactors play a very important role in the production of certain isotopes. In tracing the data obtained from operational reports related to irradiation requests from reactor users, namely PTRR, PSTNT, and PT INUKI for radioisotope production, which has been carried out in the last 5 years, May 2015 until 25 August 2020, show that the irradiation request at RSG-GAS is still not optimal. In term of the utilization of RSG-GAS, it can still be optimized, which in this case needs to be balanced with post-irradiation processing capabilities. Meanwhile, from the results of tracing and data collection, it can be shown that at this time the reactors are still operating. The utilization activities of the reactors complement each other according to their age and facilities.

STUDY ON THERMAL CHARACTERISTICS OF U-SHAPED HEAT PIPE

The latest accident in Japan's nuclear power station became a valuable experience to start engaging passive cooling systems (PCS) more aggressively to improve safety aspects in nuclear power reactors being studied in Indonesia. This investigation is related to the U-shaped heat pipe (UHP) research as PCS of water in the cooling tank (CT). The objective of this research is to study the thermal characteristics of UHP as PCS in the CT. The experiment on small-scale UHP and simulation with RELAP5 code has been conducted to understand the performance of UHP. The experiment results of the small-scale UHP model will be used as a basic understanding of simulating and designing a UHP with big scaling. The study result showed the highest thermal performance of UHP was obtained when it operated on the higher temperature of heat load and higher air cooling velocity. The more UHPs inserted into the cooling pool, the more heat that can be discharged into the environment. This result also shows promising use of UHP for CT PCS. The use of UHP as PCS can enhance the safety aspect of the nuclear reactor, especially in station blackout event.

CRITICALITY SAFETY ANALYSIS OF THE DRY CASK DESIGN WITH AIR GAPS FOR RDNK SPENT PEBBLE FUELS STORAGECriticality Safety Analysis of the Dry Cask Design with Air Gaps for RDNK Spent Pebble Fuels Storage

Reaktor Daya Non-Komersial (RDNK) with a 10 MW thermal power has been proposed as one of the technology options for the first nuclear power plant program in Indonesia. The reactor is a High Temperature Gas-Cooled Reactor-type with spherical fuel elements called pebbles. To support this program, it is necessary to prepare dry cask to safely store the spent pebble fuels that will be generated by the RDNK. The dry cask design has been proposed based on the Castor THTR/AVR but modified with air gaps to facilitate decay heat removal. The objective of this study is to evaluate criticality safety through keff value of the proposed dry cask design for the RDNK spent fuel. The keff values were calculated using MCNP5 program for the dry cask with 25, 50, 75, and 100% of canister capacity. The values were calculated for dry casks with and without air gaps in normal, submerged, tumbled, and both tumbled and submerged conditions. The results of calculated keff values for the dry cask with air gaps at 100% of canister capacity from the former to the latter conditions were 0.127, 0.539, 0.123, and 0.539, respectively. These keff values were smaller than the criticality threshold value of 0.95. Therefore, it can be concluded that the dry cask with air gaps design comply the criticality safety criteria in the aforementioned conditions.

ANALYSES OF NEUTRON ABSORBER MATERIALS ON THE SAFETY PARAMETERS IN THE RSG-GAS REACTOR

Shutdown system in RSG-GAS reactor is using neutron absorber. There are 3 kinds of absorber material in research reactors including Ag-In-Cd alloy, B4C, and Hf. In this works, analyses of different neutron absorbers on the main safety core parameters in the RSG-GAS research reactor are selected for analyses. Their integral effects on the main neutronic core parameters important to safety issues are investigated. These parameters are core excess reactivity, shutdown margin, total reactivity worth of control rods, PPF and neutron flux . The RSG-GAS core silicide fuel is selected as the case study to verify calculations. A three-dimensional, four-group diffusion model is selected for core calculations. The well-known WIMSD-5B and Batan-3DIFF reactor codes are used to carry out these calculations. It is found that the largest shutdown margin is gained using the B4C; also the lowest PPF is gained using the Hf material. The maximum point power densities belong to the inside fuel regions surrounding the CIP (centre irradiation position), surrounded by control fuel elements, and the peripheral fuel elements beside the berrylium reflector. The greatest and least fluctuation of the point power densities are gained by using B4C and Ag-In-Cd alloy, respectively.

CALCULATION OF RADIOACTIVE SOURCE TERM RELEASE FROM FLEXBLUE NPP

One of the National Research Programs (PRN) in the energy sector of the Indonesian Ministry of Research and Technology for the period of 2020-2024 is small modular reactor (SMR) nuclear power plant (NPP) assessment. The France’s Flexblue is a PWR-based SMR submerged reactor with a power of 160 MWe. The Flexblue reactor module was built on the ocean site and easily provided the supply of reactor modules, in accordance with the conditions of Indonesia as an archipelagic country. Therefore, it is necessary to know the release of fission products (source term), which is necessary for the study of the radiation safety of a nuclear reactor. This paper aims to examine the source term in normal operating conditions and abnormal normal operating conditions, as well as postulated accidents. Based on the Flexblue reactor core parameter data, the calculation of the reactor core inventory uses the ORIGEN2 software is previously evaluated. The source term calculation uses a mechanistic approach and a graded approach. The normal source term is calculated assuming the presence of impurities on the fuel plate, due to fabrication limitations. Meanwhile, the abnormal source term is postulated in the LOCA event. The core reactor inventory and source term is divided into 8 radionuclide groups which are Noble gasses group (Xe, Kr); Halogen (I); Akali Metal (Cs, Rb); Tellurium Group (Te, Sb, Sc); Barium-Strontium Group (Ba, Sr); Noble Metals (Ru, Rh, Pd, Mo, Tc, Co); Lanthanides group (La, Zr, Nd, Eu, Nb, Pm, Pr, Sm, Y, Cm, Am) and Cerium Group (Ce, Pu , Np).

ENVIRONMENTAL CONSEQUENCES OF ROUTINE RELEASES FROM SMALL MEDIUM REACTOR AT BABEL SITE

Radiation protection and safety documents for routine conditions are required to support the licensing requirements for nuclear power plant site. This research is focused in the assessment and analysis of the results of PWR safety study related to the routine release of radioactivity from the SMR subsystems and components of the 100 MWe-type PWR along with its consequences in the site. The core inventory calculation was done using ORIGEN2 software, applying release parameters from the existing analysis and calculation results. The radiological consequences were calculated by the PC-CREAM program package. Environmental and meteorological data were obtained using Arc-GIS and spatial analysis. The Bangka Belitung (Babel) site was used as the specific footprint. Analyzing PC-CREAM output data the radiological consequences of routine operation of 3 100 MWe PWR modules on Sebagin site (South Bangka) and Muntok site (West Bangka) in 16 sectors and within a radius of 20 km were concluded. The calculation results for the Sebagin site is that the maximumdose within a radius of 500 m (exclusion zone) is 1.15E+02 µSv/year. For a radius beyond 500 m, the maximum dose is 4.71E+01 µSv/year. Whereas for Muntok site (West Bangka), the maximum dose in the exclusion area (<500m) is 9.47E+00 µSv/year, and outside exclusion area (>500m) is 3.10E+00 µSv/year. The individual dose for the Babel site in the exclusion area is below the dose constraint for non-radiation service workers as the general public of 0.3 mSv/year or 300 µSv/year, while the maximum dose for outside exclusion is also below the constraint as stipulated in BAPETEN Regulation No 4 Year 2013 on Radiation Protection and Safety.

PRELIMINARY ASSESSMENT OF ENGINEERED SAFETY FEATURES AGAINST STATION BLACKOUT IN SELECTED PWR MODELS

The 2011 Fukushima accident did not prevent countries to construct new nuclear power plants (NPPs) as part of the electricity generation system. Based on the IAEA database, there are a total of 44 units of PWR type NPPs whose constructions are started after 2011. To assess the technology of engineered safety features (ESFs) of the newly constructed PWRs, a study has been conducted as described in this paper, especially in facing the station blackout (SBO) event. It is expected from this study that there are a number of PWR models that can be considered to be constructed in Indonesia from the year of 2020. The scope of the study is PWRs with a limited capacity from 900 to 1100 MWe constructed and operated after 2011 and small-modular type of reactors (SMRs) with the status of at least under licensing. Based on the ESFs design assessment, the passive core decay heat removal has been applied in the most PWR models, which is typically using steam condensing inside heat exchanger within a water tank or by air cooling. From the selected PWR models, the CPR-1000, HPR-1000, AP-1000, and VVER-1000, 1200, 1300 series have the capability to remove the core decay heat passively. The most innovative passive RHR of AP-1000 and the longest passive RHR time period using air cooling in several VVER models are preferred. From the selected SMR designs, the NuScale design and RITM-200 possess more advantages compared to the ACP-100, CAREM-25, and SMART. NuScale represents the model with full-power natural circulation and RITM-200 with forced circulation. NuScale has the longest time period for passive RHR as claimed by the vendor, however the design is still under licensing process. The RITM-200 reactor has a combination of passive air and water-cooling of the heat exchanger and is already under construction.

RADIATION DOSE OPTIMIZATION OF BREAST CANCER WITH PROTON THERAPY METHOD USING PARTICLE AND HEAVY ION TRANSPORT CODE SYSTEM

Radiotherapy is one of the cancer treatments conducted by giving a high dose to the tumor target and minimizing the dose exposed in the healthy organs. One of the methods is proton therapy. Proton therapy is usually used in several breast cancer cases by minimizing the damage in the surrounding tissues due to having good precision. In this study, proton therapy in breast cancer will be simulated. This study aims to identify the optimal dose in breast cancer therapy using proton therapy and to identify the dose exposed in the healthy organs surrounding cancer. This study is PHITS program simulation-based to model the geometry and the components of breast cancer and the surrounding organs. The source of radiation used is proton which is the output of proton therapy with proton/sec firing intensity. The variation in beam modelling towards the dose profile of the tumor used is uniform and pencil beam. The proton energy used is 70 MeV up to 120 MeV. The result of this study shows that the dose from using pencil beam scanning technic of proton therapy for breast cancer is 50.3997 Gy (W) with the total amount of fraction is 25 and the result of dose below the threshold dose in the healthy organs is the skin gets 4.4.0553 Gy per fraction, the left breast gets 0,0011 Gy per fraction, the right breast gets 2.6469 Gy per fractions, the right lung gets 0.0125 Gy per fraction, the left lung gets 0.029 Gy per fraction, the rib gets 0.0179 Gy per fraction, and the heart gets 0.0077 Gy per fraction.