Garuda - Garba Rujukan Digital

p-Index From 2020 - 2025

0.408

P-Index

This Author published in this journals

All Journal SIGMA EPSILON - Majalah Ilmiah Teknologi Keselamatan Nuklir Jurnal Teknologi Reaktor Nuklir Tri Dasa Mega Jurnal Sains dan Teknologi Nuklir Indonesia (Indonesian Journal of Nuclear Science and Technology)

Surip Widodo

Unknown Affiliation

Author-ID : 2788882

Biochemistry, Genetics & Molecular Biology Computer Science & IT Energy Environmental Science Materials Science & Nanotechnology Mechanical Engineering Medicine & Pharmacology Physics

Published : 5 Documents Claim Missing Document

Claim Missing Document

Articles

ANALYSIS OF THE EFFECT OF ELEVATION DIFFERENCE BETWEEN HEATER AND COOLER POSITION IN THE FASSIP-01 TEST LOOP USING RELAP5 Andi Sofrany Ekariansyah; Hendro Tjahjono; Mulya Juarsa; Surip Widodo
SIGMA EPSILON - Buletin Ilmiah Teknologi Keselamatan Reaktor Nuklir Vol 19, No 1 (2015): Februari 2015
Publisher : Badan Tenaga Nuklir Nasional

To understand the natural circulation phenomena on the passive residual heat removal system (PRHRS), development of a test section describing that phenomena in particular in the one phase condition is required. That test facility is named as FASSIP-01 in form of a vertically closed loop consisting of piping compo-nents, one cylinder tank featured with heater elements and one cooler. The heater tank will work as the heat source, and the cooler as the heat sink. This research is intended to support the experimental activity of the FASSIP-01 by conducting a simulation using the RELAP5/SCDAP/Mod3.4. Beside the standard loop configuration, the simulation is also conducted by varying the elevation of heater and cooler position to evaluate the best position resulting in the most optimal natural circulation. The results will be used as the comparison with the later performed experiment. The simulation result shows that for the case where the heater position is at the same level with the cooler position, the temperature distribution of the water after the heater and after the cooler are higher than the other two position. Looking at the natural circulation, that position results in the lowest mass flow. The position with the heater below the cooler will result in the best mass flow. On that position, only an optimiza-tion in the heat transfer surface area is needed to increase the heat transfer coefficient and secondary mass flow to remove the heat are needed to obtain more optimal performance of the water circulation caused by the density difference in the FASSIP-01 test loop.

PRELIMINARY DEVELOPMENT OF RADIONUCLIDES RELEASE OF INDIVIDUAL DOSE CODE PROGRAM FOR RADIATION MONITORING PURPOSES Jupiter Sitorus Pane; Pande Made Udiyani; Muhammad Budi Setiawan; Surip Widodo; I Putu Susila
JURNAL TEKNOLOGI REAKTOR NUKLIR TRI DASA MEGA Vol 23, No 3 (2021): October (2021)
Publisher : Pusat Teknologi Dan Keselamatan Reaktor Nuklir (PTKRN)

Show Abstract | Download Original | Original Source | Check in Google Scholar | DOI: 10.17146/tdm.2021.23.3.6240

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.

DEVELOPMENT OF EXPERIMENTAL POWER REACTOR (EPR) MODEL FOR SAFETY ANALYSES USING RELAP5 Andi Sofrany Ekariansyah; Muhammad Subekti; Surip Widodo; Hendro Tjahjono; Susyadi Susyadi; Puradwi Ismu Wahyono; Anwar Budianto
JURNAL TEKNOLOGI REAKTOR NUKLIR TRI DASA MEGA Vol 21, No 2 (2019): JUNI 2019
Publisher : Pusat Teknologi Dan Keselamatan Reaktor Nuklir (PTKRN)

Pebble bed reactor design, classified as the high temperature gas-cooled reactor (HTGR), is currently being part of BATAN main program to promote nuclear energy by starting the Experimental Power Reactor (EPR) program since 2015. Starting from 2018, the detail design document has to be submitted into nuclear regulatory body for further assessment. Therefore results of design analysis have to be supplemented by performing a design evaluation, which can be achieved by developing the model of the EPR. The development is performed using RELAP5/SCDAP/Mod.3.4 as the thermal-hydraulic analysis code validated for the light-water reactor having module for the pebble fuel element and non-condensable helium gas. Methodology of model development consists of defining the helium flow path inside the reactor pressure vessel, modelling of pebble bed core including its power distribution, and modelling of reflector components to be simulated under 100 % core power. The developed EPR model results in design parameters, which confirm the main thermal data of the EPR, including the pebble and reflector temperatures. The peak pebble temperature is calculated to be 1,375 °C, which requires further investigations in the model accuracy, since the reference values are around 1,015 °C, even it is below the pebble temperature limit. For safety analysis, the EPR model can be used under nominal core flow condition, which produces more conservative results by paying attention on the RELAP5 specific modules for the pebble bed-gas cooled system.Keywords: experimental power reactor, development, RELAP5, steady-state

PRELIMINARY ASSESSMENT OF ENGINEERED SAFETY FEATURES AGAINST STATION BLACKOUT IN SELECTED PWR MODELS Andi Sofrany Ekariansyah; Surip Widodo; Susyadi Susyadi; Hendro Tjahjono
JURNAL TEKNOLOGI REAKTOR NUKLIR TRI DASA MEGA Vol 23, No 2 (2021): June 2021
Publisher : Pusat Teknologi Dan Keselamatan Reaktor Nuklir (PTKRN)

Show Abstract | Download Original | Original Source | Check in Google Scholar | DOI: 10.17146/tdm.2021.23.2.6204

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.

SIMULATION OF OPERATIONAL CONDITIONS OF FASSIP-02 NATURAL CIRCULATION COOLING SYSTEM EXPERIMENTAL LOOP Dr.Anhar Riza Antariksawan; Surip Widodo; Mulya Juarsa; Dedy Haryanto; Mukhsinun Hadi Kusuma; Nandy Putra
Jurnal Sains dan Teknologi Nuklir Indonesia (Indonesian Journal of Nuclear Science and Technology) Vol 19, No 1 (2018): Februari 2018
Publisher : BATAN

The natural circulation is considered in the design of emergency passive core cooling system in a nuclear power plant. In that context, in order to investigate the characteristics of the natural circulation, FASSIP-02 experimental loop is designed. This paper simulates the various operational conditions with different condition of the heater power, the pipe insulation and the expansion tank's valve using RELAP5 code. The objective is to obtain the best operational conditions of FASSIP-02 once it is built. The simulation results show that the until 50,000 s the steady state condition could not be achieved yet when the heater power greater than 10 kW. The pipe insulation reduced the heat loss to the environment and in turn it causes faster increase of the water temperature inside the pipe. While, if the expansion tank's valve is closed during the operation, the pressure inside the loop would increase, faster when the heater power is higher. It is concluded that in all cases to avoid the saturation condition, the heater power should be maintained lower than 10 kW, especially when the loop pipe is insulated.

Co-Authors Andi Sofrany Ekariansyah Andi Sofrany Ekariansyah Andi Sofrany Ekariansyah Anwar Budianto Dedy Haryanto Dr.Anhar Riza Antariksawan Hendro Tjahjono I Putu Susila Jupiter Sitorus Pane Muhammad Budi Setiawan Muhammad Subekti Mukhsinun Hadi Kusuma Mulya Juarsa Nandy Putra Pande Made Udiyani Puradwi Ismu Wahyono Susyadi Susyadi

Title

Found 5 Documents
Search

Abstract

Abstract

Abstract

Abstract

Abstract

Title Search

Found 5 Documents Search

Abstract

Abstract

Abstract

Abstract

Abstract

Title

Found 5 Documents
Search