Accident Source Term Research Articles

Source term inversion of nuclear accidents based on ISAO-SAELM model

The release source term of radioactivity becomes a critical foundation for emergency response and accident consequence assessment after a nuclear accident Rapidly and accurately inverting the source term remains an urgent scientific challenge. Today source term inversion based on meteorological data and gamma dose rate measurements is a common method. But gamma dose rate actually includes all nuclides information, and the composition of radioactive nuclides is generally uncertain. This paper introduces a novel nuclear accident source term inversion model, which is Improve Snow Ablation Optimizer-Sensitivity Analysis Pruning Extreme Learning Machine (ISAO-SAELM) model. The model inverts the release rates of 11 radioactive nuclides (I-131, Xe-133, Cs-137, Kr-88, Sr-91, Te-132, Mo-99, Ba-140, La-140, Ce-144, Sb-129). It does not require the use of the physical field of the reactor to obtain prior information and establish a dispersion model. And the robustness is validated through noise analysis test. The mean absolute errors of the release rates of 11 nuclides are 15.52 %, 15.28 %, 15.70 %, 14.99 %, 14.85 %, 15.61 %, 15.96 %, 15.42 %, 15.84 %, 15.13 %, 17.72 %, which show the significant superiority of ISAO-SAELM. ISAO-SAELM model not only achieves notable advancements in accuracy but also receives validation in terms of practicality and feasibility.

Research of aerosol removal by passive containment cooling system

Passive containment cooling system (PCCS) can effectively remove the heat from the containment and maintain the pressure and temperature in case of accident, thus ensuring the containment integrity. PCCS introduces thermophoresis and diffusiophoresis aerosol deposition and changes the distribution of fission products in the containment, thus affecting containment dose evaluation and off-site source term assessment. In this paper, a Generation III PWR nuclear power plant model with PCCS is established by using ASTEC (Accident Source Term Evaluation Code), and the removal effect of PCCS on fission product aerosol is studied under the large break loss of coolant accident. The results show that one train of PCCS can increase the aerosol removal factor for approximately 0.05/h. Taking five trains of PCCS in this study as an example, aerosol deposition effect introduced by PCCS reduces suspended aerosol in containment by 10 %–20 %, and the effect is more significant in the PCCS-locating compartment. Meanwhile, the gas flow among compartments changes the aerosol mass in the adjacency of the PCCS-locating compartment. This phenomenon weakens the removal effect in the PCCS-locating compartment and make the aerosol concentration in adjacent compartments tend to be consistent.

Validation and uncertainty analysis of ASTEC in early degradation phase against QUENCH-06 experiment

Severe Accident (SA) integral codes, such as the Accident Source Term Evaluation Code (ASTEC) developed by IRSN, are used to simulate the phenomena occurring during accident progression in Nuclear Power Plants (NPPs) up to the source term evaluation. Code validation against experimental data is fundamental to carry out deterministic safety analysis and apply these codes to NPPs. In addition, in the Best Estimate Plus Uncertainty (BEPU) framework, the quantification of the results uncertainty is needed. In the framework of the IAEA CRP I31033 “Advancing the State-of-Practice in Uncertainty and Sensitivity Methodologies for Severe Accident Analysis in Water-Cooled Reactors”, the QUENCH test-6 experiment, conducted at KIT, has been selected to develop an uncertainty analysis using the ASTEC v2.2b code. The accuracy of the best-estimate ASTEC simulation was evaluated with the Fast Fourier Transform Based Method (FFTBM) against the experimental data. Then, the uncertainty of the code results was quantified by using the probabilistic propagation of input uncertainties method, through the coupling of ASTEC with RAVEN (Risk Analysis and Virtual Environment). Beyond identifying the main sources of uncertainty affecting the simulated test, the outcomes of the work also include some general discussion on the uncertainty propagation in a SA sequence.

Accident source term and radiological consequences of a small modular reactor

Accident source term and radiological consequences of a small modular reactor

Influence of guide tubes on configuration and heat-up behaviour for PHWR specific suspended debris bed – An analytical assessment

The estimation of accident source term and hydrogen generation is dependent on the core heat-up under postulated accident scenarios. The uncertainty of such estimations increases as the accident severity increases due to complexity in the determination of i) evolving degraded core configurations, and ii) associated suitable physical models for heat-up and degradation simulation. Analysis of postulated severe accidents for PHWR is essential for emergency planning and severe accident management guidelines from a plant licensing perspective. The study presents the order of uncertainty that is present as of today in deciding the size and shape of the suspended debris bed, which in turn decides the exposed core heat up, hydrogen generation as well as fission product release. The study reveals that the location and integrity of reactivity mechanism guide tubes significantly impact the exposed core heat-up, channel disassembly pattern and shape and size of the suspended debris bed. Nevertheless, the experimental study on the guide tube behaviour on asymmetric heat-up condition and their failure will further substantiate the postulation of suspended debris bed configuration.

Application of the MOCABA algorithm and the FSTC tool for source term predictions during severe accident scenarios

This paper describes the capabilities of the Fast Source Term Calculation (FSTC) tool developed by the Karlsruhe Institute of Technology (KIT) to perform time series predictions of the radiological source term (ST) during severe accident (SA) scenarios. The prediction algorithm implemented in FSTC is based on the MOCABA data assimilation framework, developed by Framatome, and requires a training database and actual measurements. The training database is assessed by employing coupling FSTC to the European reference Accident Source Term Evaluation Code (ASTEC), developed by IRSN. The prediction capabilities of FSTC is applied to evaluate: 1) the mass of released hydrogen during the QUENCH-08 experiment; 2) the xenon release to the environment during a Medium Break Loss of Coolant Accident (MBLOCA) SA scenario in a generic KONVOI nuclear power plant (NPP). The prediction results show a very promising employment of FSTC and the MOCABA algorithm in view of supporting the emergency response team during SA scenarios.

MELCOR 2.2-ASTEC V2.2 crosswalk study reproducing SBLOCA and CSBO scenarios in a PWR1000-like reactor part II: Analysis of containment thermal-hydraulics and ex-vessel phenomena

The Accident Source Term Evaluation Code (ASTEC) and Methods of Estimation of Leakages and Consequences of Releases (MELCOR) system codes are both used by Tractebel to perform Severe Accident (SA) analyses for the Belgian Nuclear Power Plants (NPPs).This paper contains the second part of the comparative study between these two integral software focused on the SA progression in the containment after the Vessel Failure (VF).The main phenomena characterize this phase of the accident usually called ex-vessel phase are the Molten Corium Concrete Interaction (MCCI) and ex-vessel corium/debris coolability, which determine the containment Thermal-Hydraulics (TH) response and can cause its late failure. Despite their importance from a safety point of view, chemical and physical processes behind these phenomena are not yet well understood, partly because of their extreme complexity and partly because of the limitations of current experimental capabilities in reproducing them. As result, models developed and implemented in the codes to describe the MCCI and debris coolability are extremely simplified because based on limited data availability.The benchmark study consists of a comparison of the results calculated by the ASTEC and MELCOR codes during the ex-vessel phase following two different scenarios a Complete Station Blackout (CSBO) and a Small Break Loss Of Coolant Accident (SBLOCA).For each scenario, two different Severe Accident Management Strategies (SAMS) are adopted, the first one is to ensure that the corium, once fully relocated in the cavity, has top-flooding conditions throughout the transient, while the second one is to keep the reactor cavity completely dry. Furthermore, for every case reproduced the reference and best-practice models are investigated.The results show that in case of dry cavity conditions, despite the different VF timing calculated by the two codes, the final results provide very similar indications even though some dissimilarity is observed regarding the amount of fission product (FP) released through the Containment Filtering Venting System (CFVS).In contrast, in the case of wet cavity conditions, the accident progression predicted is significantly different regarding number of venting operations, TH containment conditions, and amount of concrete ablated.The discrepancies observed, should be sought in the corium and debris coolability models, which given their parametric nature and oversimplification can lead to these results.

DETERMINATION OF POTENTIAL AREAS OF INFLUENCE ON THE ROMANIAN TERRITORY OF UKRAINIAN NUCLEAR POWER PLANTS IN THE CONTEXT OF THE CRISIS GENERATED BY THE RUSSIAN-UKRAINIAN CONFLICT

In the context of the crisis generated by the Russian-Ukrainian conflict, radiological situation forecasts were made within the National Institute of Research and Development for Physics and Nuclear Engineering "Horia Hulubei" (IFIN-HH) for all nuclear power plants in Ukraine. Therefore, the resident decision support software systems for radiological and nuclear emergencies were used, with the main purpose of identifying potential areas of influence on the territory of Romania generated by Ukrainian power plants, as consequence of military actions. In the absence of pertinent and confirmed information, two hypothetical scenarios were considered by which the radioactive material could be dispersed into the atmosphere, for which appropriate accident source terms were built, using methods validated and approved by the international organizations in the field. The monitoring of the radiological situation of the Ukrainian nuclear power plants was carried out over a period of several months starting from the first day of the conflict outbreak. In the paper are presented comparative results obtained with two software systems for a representative case at the Zaporozhe Nuclear Power Plant, the largest nuclear power plant in Europe, situated in the center of several disputes.

Design and simulation of a counter measure analysis system in a decision support system ONERS for nuclear emergency

In this work, a Counter Measures Analysis System (CMAS) is developed in the Online Nuclear Emergency Response System (ONERS) to suggest protective actions for minimizing radiation exposure to people during accidental releases of radionuclides to the environment. CMAS is designed with PHP, Java and MySQL database engine in Web-GIS environment to enable online computations and visualization of results. CMAS maps the dispersion model simulated radiological doses to each village unit by spatial aggregation of dose from all the grid cells contained within village unit and delineates the domain into areas assigned with several protective actions taking into account of dose criteria. It also computes residual and averted doses along with timing of different protective actions in each village. The effectiveness of CMAS is tested by simulating the consequences for a hypothetical accident. Results suggest that CMAS provides realistic estimates of protective actions following the spatial dispersion and dose pattern at each village unit. Comparison with conventional grid-based calculations indicate CMAS provides more practical suggestion of protection strategy compared to grid-based suggestions due to the conservative nature of estimates and attribution of all necessary information for emergency management. Sensitivity tests with respect to accident source term and dose criteria indicate that CMAS produces corresponding variation in impacted areas and protective actions confirming consistent performance.

Uncertainty and sensitivity analysis of the QUENCH-08 experiment using the FSTC tool

This paper presents results of a preliminary uncertainty and sensitivity (U&S) analysis of the QUENCH-08 experiment using the Fast Source Term Calculation (FSTC) tool, which currently is under development at KIT/INR as a part of WAME project. WAME is devoted to the development of a methodology and a tool for fast and accurate source term (ST) calculation in case of severe accident at nuclear power plant (NPP). The FSTC tool is divided into 2 parts: the U&S analysis and ST prediction (using the MOCABA approach), respectively. The description of the MOCABA approach, which is implemented in FSTC tool, will be presented – together with the results of ST evaluation – in a subsequent paper. In this paper, the FSTC tool is applied in conjunction with the Accident Source Term Evaluation Code (ASTEC) (version 2.2_b) severe accident (SA) code to the U&S analysis of the QUENCH-08 experiment to demonstrate the functionality of the tool and to check the methodology’s suitability for the application to a nuclear power plant.

Reassessment of Computational Tools for the Modeling of Heat Transfer in a Molten UO2 Pool in Natural Convection

Abstract This work, performed within the European sodium fast reactor safety measures assessment and research tools (ESFR-SMART) H2020 European project, is part of a larger framework intending to reassess the modeling of heat transfer in molten pools on SCARABEE available experimental results. This paper presents simulation results of the in-pile Bain Fondu (BF1) test, performed within the SCARABEE-N program, using accident source term evaluation code (ASTEC), Sn method implicit multifiel multiphase Eulerien recriticality (SIMMER) III, and SIMMER V simulation tools as well as comparison with its available experimental data. This program was performed in the 1980s in the frame of the Safety Assessment studies of Superphenix sodium-cooled reactor. This test was dedicated to verify the stability of a pure molten UO2 pool under decay heat conditions within natural convection and the long-term resilience of the peripheral fuel crust. The pool was generated in a stainless steel crucible by a progressive heating of a fuel pellet stack through six successive power plateaus. For the benchmark purposes, only the molten pool behavior k the last power plateau, where the pool was the largest and the fuel temperatures the highest, was investigated. Experimental data such as the axial profile of radial heat fluxes and heat transfer from the pool to the surrounding interassembly coolant as well as the axial profile of the peripheral fuel crust thickness were used for the reassessment of the simulation tools. In addition, other variables of interest not measured during the test, such as the radial and axial velocities in the pool, were also benchmarked. Finally, a critical analysis of the correlations and models used in the different simulation tools for the BF1 test modeling is also provided in the paper.

Severe accident simulation for VVER-1000 reactor using ASTEC-V2.1.1.3

Abstract The station black-out (SBO) is one of the main accident sequences to be considered in the field of severe accident research. To evaluate a nuclear power plant’s behavior in the context of this accident, the integral ASTEC-V2.1.1.3 code “Accident Source Term Evaluation Code” covers sequences of SBO accidents that may lead to a severe accident. The aim of this work is to discuss the modelling principles for the core melting and in-vessel melt relocation phenomena of the VVER-1000 reactor. The scenario of SBO is simulated by ASTEC code using its basic modules. Then, the simulation is performed again by the same code after adding and activating the modules; ISODOP, DOSE, CORIUM, and RCSMESH to simulate the ex-vessel melt. The results of the two simulations are compared. As a result of SBO, the active safety systems are not available and have not been able to perform their safety functions that maintain the safety requirements to ensure a secure operation of the nuclear power plant. As a result, the safety requirements will be violated causing the core to heat-up. Moreover potential core degradation will occur. The present study focuses on the reactor pressure vessel failure and relocation of corium into the containment. It also discusses the transfer of Fission Products (FPs) from the reactor to the containment, the time for core heat-up, hydrogen production and the amount of corium at the lower plenum reactor pressure vessel is determined.

Analysis of Sentitivity of Fission Product Iodine in Containment to Various Factors under Severe Accidents

The most serious consequence of reactor accident is that the radioactive fission products are dispersed into the environment. To study the distribution characteristics of radioactive fission product iodine in the containment under severe accident conditions, this paper assumes that the fission product iodine is released from the primary system into the containment due to the occurrence of severe accident in the nuclear power plant (NPP). Then, using the accident source term evaluation code (ASTEC), this study builds a containment construction model for the NPP, sets the boundary conditions, and calculates the chemical form, chemical properties, distribution, and change trends of different compounds, of the fission product iodine under different pH values, with or without silver (Ag) injection and under gaseous radiation conditions. The results show that the production of volatile iodine in the containment can be inhibited under alkaline conditions; the silver can promote the iodine trap in the liquid phase and reduce the iodine volatility; and the gaseous radiation environment can promote the formation of gaseous CH3I and IOx. As a result, this study can provide guidance for the removal of radioactive iodine in the containment under serious accidents.

MELCOR 2.2-ASTEC V2.2 crosswalk study reproducing SBLOCA and CSBO scenarios in a PWR1000-like reactor part I: Analysis of RCS thermal-hydraulics and in-vessel phenomena

The Accident Source Term Evaluation Code (ASTEC) and Methods of Estimation of Leakages and Consequences of Releases (MELCOR) system codes are used in Tractebel to perform source term assessments and plant Thermal Hydraulic (TH) analyses in case of Severe Accident (SA) for the Belgian Nuclear Power Plants (NPPs).These system codes were developed through intense and extensive validation activities which were mainly based on the verification of the capability of implemented physical models to reproduce the experimental results of Separate Effect Tests (SETs) and Coupled Effect Tests (CETs).However, the SETs, and in less extension the CETs, do not take into account interactions and possible synergistic or antagonistic effects which can arise among the different physical and chemical processes occurring during a SA. Thus, if a code is able to reproduce correctly most of SETs or CETs, this does not mean that it can ensure the same quality of results for the more realistic and complex Integral Effect Tests (IETs).Nevertheless, some contraindications exist also for the IETs, because of the scaling factor effect which, being difficult to quantify, could lead to development of models replicating correctly the experimental data but that are not totally reliable for reactor scale applications.In addition, waiting for the dismantling of Fukushima Dai-ichi units 1–2 and 3, which can provide new information in the field of the core degradation phenomena, in the present only data on in-vessel core melt progression for the Three Mile Island, Unit 2 (TMI-2) accident are available.Therefore, apart from simulating the TMI-2 accident scenario, there is no other way to assess models of in-vessel phenomena (H2 production, corium pool/ debris formation and relocation, etc) implemented in the codes on reactor scale.As a consequence, there are still uncertainties in the capability of these integral tools to predict the progression of postulated severe accident transients in a NPP, notably core degradation and ex-vessel phenomena.In order to investigate the differences among the physical models adopted in MELCOR 2.2 and ASTEC V2.2 codes, and evaluate the impact that they could have on the assessment of Severe Accident Management Strategies, a crosswalk study was carried-out.This paper describes the first phase of this work, which consists of a detailed comparison between these two SA tools on a well-defined plant (PWR1000-Like) with prescribed boundary and initial conditions.Furthermore, to avoid additional and unwanted sources of discrepancies between code calculations, the two reactor models were developed in parallel following the recommendations suggested by developers for core and Reactor Coolant System (RCS) nodalization.The comparative study is focused on the TH of the RCS and in-vessel phenomena, for two different accident scenarios (SBLOCA and CSBO) up to the moment of lower-head failure.Ex-vessel behaviour will be examined in the second phase of this study.The crosswalk results obtained have shown some differences and similarities on reproducing TH in the RCS and In-vessel phenomena.Especially for these latter, the discrepancies predominate mainly due to how the codes treat corium behaviour, while thanks to the harmonisation of the initial steady-state and boundary conditions, the discrepancies on the prediction of the RCS behaviour have been minimized, at least before significant core degradation takes place.

Rapid accident source term estimation (RASTE) for nuclear emergency response in high temperature gas cooled reactor

Rapid accident source term estimation (RASTE) is one of the key functions required by nuclear power plant (NPP) emergency. The results of RASTE are provided to the nuclear decision-making system in emergency, response to ranking accident by INES table and determining suitable protection. This paper proposes the RASTE system for high temperature gas-cooled reactor (HTR) to estimate accident source term based on the simplified methodology used in safety analysis report (SAR) and more realistic operation condition before and in an accident. RASTE simplifies the expression for normal radioactivity in the primary loop, re-suspended radioactivity, additional radioactivity release during the heating process, graphite dust generation, desorbed radioactivity from corrosion of graphite matrix and radioactivity released from failed coated fuel particle. Modifications are made to correct the different ingress water amounts and different fuel burnup. Some unclearly known parameters and processes are estimated from the operation experience as well as accident analysis by conservation rule. The integrated code of fundamental framework and user interface are also developed. This paper takes the High Temperature Gas Cooled Reactor Pebble-bed Module (HTR-PM), a demonstration project being constructed at Shidao Bay, Shandong Province, China, as the study case to discuss the applicability of RASTE. Considering three typical HTR-PM design basis accidents (DBAs), this paper selects P2, P4 as the representatives of LOCA accidents, selects SG6 and SG91 as the representatives of ingress accidents and selects T2 as the representative of transient accidents. Accuracy verification is made by comparing the environmental release estimated by RASTE to that in SAR. These compared targets include nuclide groups of inert gas, iodine, and aerosols, and key nuclides of Xe-133, I-131, Cs-134. Calculation results show that the methodology presented and RASTE is fit to estimate accident of HTR-PM. The development of RASTE system will contribute to form adequate capability of rapid emergency response for HTR nuclear power plant.

SOURCE TERM ASSESSMENT FOR 100 MWE PRESSURIZED WATER REACTOR

One of the barriers on the implementation of nuclear energy in Indonesia is public perception towards the safety of nuclear power plants (NPPs). Therefore, it is necessary to perform a study about the radiation impact of normal and abnormal operations of an NPP. In accordance to the program of Ministry of Research and Technology period 2020-2024, concerning the plan to build a small modular reactor (SMR)-type NPP, a radiation safety study has been performed for the 100 MWe Pressurized Water Reactor (PWR-100MWe). Source term release of radioactive substances into the environment from PWR-100MWe is a starting point in the study of the radiological consequences of reactor operation. Therefore, this paper will examine the PWR-100MWe source term under normal and abnormal operating conditions, according to the design and the design basis accident (DBA). The initial trigger of the DBA is Lost of Coolant Accident (LOCA) such as Small LOCA and Large LOCA. Due to the limitations of available SMR data, the study of PWR-100MWe source term refers to the assumption of the release fraction of fission products per subsystem in a larger 1000MWe PWR. It is expected from this assumption that pessimistic source term will be obtained. The study begins with calculation of PWR-100MWe core inventory using ORIGEN2 code based on PWR-100MWe reactor parameters. Through the mechanistic source term model and PWR-1000MWe release parameters, source terms will be obtained for normal operation and abnormal conditions i.e. DBA. Normal source term is used to calculate the consequences of normal operation, which will be used for environmental monitoring and environmental safety analysis of the site. Whereas accident source term is the basis for calculating the radiological consequences of accidents used for SAR documents and nuclear preparedness.Keywords: SMR, PWR-100MWe, normal operation, source term, accident

Uncertainty & sensitivity analyses for LOCA simulation of PHEBUS experiment FPT-1 using a modified SUNSET code

One task of the regulatory body during the review and assessment process is the quantification of uncertainty and sensitivity (U&S) during normal and accident conditions. Nowadays there are many codes having the capability of estimating U&S, but few of them can estimate uncertainty and sensitivity during severe accident conditions. The IRSN uncertainty propagation and sensitivity analysis tool named SUNSET is considered one of the most important tools used in U&S analysis, where it is coupled with ASTEC software package (Accident Source Term Evaluation Code) for computing U&S during severe accident condition. Unfortunately, the SUNSET tool cannot perform U&S analyses during all the transient time. The SUNSET software can only calculate U&S parameters for maximum, minimum, and last output values, therefore it was necessary to modify the SUNSET tool to calculate U&S during the whole transient time. The main objective of this study is to develop a Python script to replace the SUNSET post-processing step, extract the results, and perform U&S analysis. ASTEC code has been used for the simulation of a loss of coolant accident in the cold leg of the primary circuit of the FPT-1 test. The obtained results indicate the efficiency of the modified SUNSET tools to calculate the uncertainty during severe accidents. In general, it was found that the code is capable to model the main severe accident phenomena. SUNSET package shows good agreement with the experimental data, for the whole calculation period, until thermocouples failure moment. The obtained results show that the main safety parameters are very sensitive to the power multiplier.

Validation of Selected Cesar Friction Models of the ASTECV21 Code Based on Moby Dick Experiments

In the event of a loss of integrity of the main coolant line, a large mass and energy release from the primary circuit to the containment is to be expected. The temporal evolution of such depressurization is mainly governed by the critical flow, whose correct prediction requires, in first place, a correct description of the different friction terms. Within this work, selected friction models of the CESAR module of the Accident Source Term Evaluation Code (ASTEC) V2.1 integral code are validated against data from the Moby Dick test facility. Simulations are launched using two different numerical schemes: on the one hand, the classical five equation (drift flux) approach, with one momentum conservation equation for an average fluid plus one algebraic equation on the drift between the gas and the liquid; on the other hand, the recently implemented six equation approach, where two differential equations are used to obtain the phase velocities. The main findings are listed hereafter: The use of five equations provides an adequate description of the pressure loss as long as the mass fluxes remain below 1.24 kg/cm2 s and the gas mass fractions below 5.93 × 10 − 4. Beyond those conditions, the hypotheses of the drift flux model are exceeded and the use of an additional momentum equation is required. The use of an additional momentum equation leads to a better agreement with the experimental data for a wider range of mass fluxes and gas mass fractions. However, the qualitative prediction for high gas mass fractions still shows some deviations due to the decrease of the regular friction term at the end of the test section.

THE STUDY OF ATMOSPHERIC DISPERSION MODEL ON ACCIDENT SCENARIO OF RESEARCH REACTOR G. A. SIWABESSY USING HOTSPOT CODES AS A NUCLEAR EMERGENCY DECISION SUPPORT SYSTEM

G.A. Siwabessy Multipurpose Reactor (RSG-GAS) is a research reactor with thermal power of 30 MW located in the Serpong Nuclear Area (KNS), South Tangerang, Banten, Indonesia. Nuclear emergency preparedness of RSG-GAS needs to be improved by developing a decision support system for emergency response. This system covers three important aspects: accident source terms estimation, radioactive materials dispersion model into the atmosphere and radiological impact visualization. In this paper, radioactive materials dispersion during design basis accident (DBA) is modeled using HotSpot, by utilizing site-specific meteorological data. Based on the modelling, maximum effective dose and thyroid equivalent dose of 1.030 mSv and 26 mSv for the first 7 days of exposure are reached at distance of 1 km from the release point. These values are below IAEA generic criteria related to risk reduction of stochastic effects. The results of radioactive dispersion modeling and radiation dose calculations are integrated with Google Earth Pro to visualize radiological impact caused by a nuclear accident. Digital maps of demographic and land use data are overlayed on Google Earth Pro for more accurate impact estimation to take optimal emergency responses.Keywords: G.A. Siwabessy research reactor, Nuclear emergency, Atmospheric dispersion model, Decision support system, HotSpot codes

A NEW METHOD FOR NUCLEAR ACCIDENT SOURCE TERM INVERSION BASED ON GA-BPNN ALGORITHM

A NEW METHOD FOR NUCLEAR ACCIDENT SOURCE TERM INVERSION BASED ON GA-BPNN ALGORITHM