TRICO II Core Inventory Calculation and its Radiological Consequence Analyses

2016 ◽  
Vol 2 (2) ◽  
Author(s):  
J. L. Muswema ◽  
G. B. Ekoko ◽  
J. K.-K. Lobo ◽  
V. M. Lukanda ◽  
E. K. Boafo
Keyword(s):  
Fuel Element ◽  
Research Reactor ◽  
Source Term ◽  
Democratic Republic Of Congo ◽  
Reactor Core ◽  
Lessons Learned ◽  
Severe Accident ◽  
Plane Crash ◽  
Reactor Building ◽  
Accident Scenarios

Two severe accident scenarios are investigated in this paper as they have never been considered previously in the safety analysis report (SAR) of the Congo TRIGA Mark II research reactor (TRICO II) in Kinshasa, the Democratic Republic of Congo. The source term is derived from the reactor core after two postulated accidents: (1) a large plane crash with total destruction of the reactor building and (2) full damage of one fuel element while the reactor building remains intact. Total effective dose (TED), after core inventory, and dose profiles to human organs are derived to assess the operational safety of the reactor. Results from the study will be used to upgrade the current SAR of the reactor as the reactor safety and licensing concepts are changing over the years; the knowledge and lessons learned from the past experience are being updated accordingly with the available data. TEDs to workers of the facility show that higher values are obtained at areas near the source term at the time of the plane crash accident, which dies out as quickly as the plume is carried away following predominant meteorological conditions at the site. The situation with one fuel element totally damaged poses no threat as far as radiation protection is concerned and reveals a maximum TED of 1.30×10−7  mSv at 100 m from the reactor core. This shows that the operation of this type of research reactor is reliable and safe.

A Research Reactor Source Term and Dose Results in the Event of a LOCA

2014 ◽  
Author(s):  
Charalampos Pappas ◽  
Andreas Ikonomopoulos ◽  
Athanasios Sfetsos ◽  
Spyros Andronopoulos ◽  
Melpomeni Varvayanni ◽  
...  
Keyword(s):  
Research Reactor ◽  
Source Term ◽  
Dispersion Model ◽  
Ventilation System ◽  
Atmospheric Dispersion ◽  
Reactor Core ◽  
Forecast Model ◽  
Scale Model ◽  
Continuous Reactor ◽  
Reactor Building

The present study discusses the source term derivation and dose result calculation for a hypothetical accident sequence in the Greek Research Reactor – 1 (GRR-1). A loss-of-coolant accident (LOCA) has been selected as a credible accident sequence. The source term derivation has been based on the GRR-1 confinement performance where the inventory has been computed assuming continuous reactor operation. A core damage fraction of 30% has been considered for the calculations while conservative core release fractions have been employed. The radionuclides released from the reactor core to the confinement atmosphere have been subjected to natural decay, deposition on and resuspension from various internal surfaces before being led to the release pathway. It has been assumed that an emergency shutdown is initiated immediately after the beginning of the accident sequence and the emergency ventilation system is also activated. Subsequently, the source term has been derived comprising of noble gases, iodine and aerosol. The JRODOS computational software for off-site nuclear emergency management has been utilized to estimate the dose results from the LOCA-initiated source term that is released in its entirety from the reactor stack at ambient temperature. The Local Scale Model Chain in conjunction with the DIPCOT atmospheric dispersion model that is embedded in JRODOS have been used with proper parameterization of the calculation settings. Five weather scenarios have been selected as representative of typical meteorological conditions at the reactor site. The scenarios have been assessed with the use of the Weather Research and Forecast model. Total effective, skin, thyroid, lung and inhalation doses downwind of the reactor building and up to a distance of 10 km have been calculated for each weather scenario and are presented. The total effective gamma dose rate at a fixed distance from the reactor building has been assessed. The radiological consequences of the dose results are discussed.

Effects of Ruthenium Release in Oxidizing Conditions on a BWR Source Term

2009 ◽  
Author(s):  
Atso Suopaja¨rvi ◽  
Teemu Ka¨rkela¨ ◽  
Ari Auvinen ◽  
Ilona Lindholm
Keyword(s):  
Pressure Vessel ◽  
Source Term ◽  
Reactor Core ◽  
Severe Accident ◽  
Ruthenium Oxides ◽  
The Core ◽  
The Standard Model ◽  
Reactor Building ◽  
Release Model ◽  
Scoping Studies

The release of ruthenium in oxygen-rich conditions from the reactor core during a severe accident may lead to formation of significantly more volatile ruthenium oxides than produced in steam atmosphere. The effect of volatile ruthenium release in a case a reference BWR nuclear plant was studied to get rough-estimates of the effects on the spreading of airborne ruthenium inside the containment and reactor building and the fission product source term. The selected accident scenario starting during shutdown conditions with pressure vessel upper head opened was a LOCA with a break in the bottom of the RPV. The results suggest that there is a remarkable amount of airborne Ru in the containment atmosphere, unlike with the standard MELCOR Ru release model which predicts no airborne Ru at all in the selected case. The total release of ruthenium from the core can be 5000 times the release predicted by the standard model. Based on the performed plant scoping studies it seems reasonable to take the release of volatile ruthenium oxides into account when assessing source terms for plants during shutdown states.

scholarly journals Source Term Derivation and Radioactive Release Evaluation for JRTR Research Reactor under Severe Accident

2020 ◽  
Vol 2020 ◽  
pp. 1-9 ◽  
Author(s):  
Ned Xoubi
Keyword(s):  
Research Reactor ◽  
Source Term ◽  
Atmospheric Dispersion ◽  
Reactor Core ◽  
Fission Products ◽  
Severe Accident ◽  
Safety Standards ◽  
Gaussian Plume ◽  
Public Exposure ◽  
Radioactive Release

The source term for the JRTR research reactor is derived under an assumed hypothetical severe accident resulting in generation of the most severe consequences. The reactor core is modeled based on the reactor technical design specifications, and the fission products inventory is calculated by using the SCALE/TRITON depletion sequence to perform burnup and decay analyses via coupling the NEWT 2-D transport lattice code to the ORIGEN-S fuel depletion code. Fifty radioisotopes contributed to the evaluation, resulting in a source term of 3.7 × 1014 Bq. Atmospheric dispersion was evaluated using the Gaussian plume model via the HOTSPOT code. The plume centerline total effective dose (TED) was found to exceed the IAEA limits for occupational exposure of 0.02 Sv; the results showed that the maximum dose is 200 Sv within 200 m from the reactor, under all the weather stability classes, after which it starts to decrease with distance, reaching 0.1 Sv at 1 km from the reactor. The radiation dose plume centerlines continue to the exceed international basic safety standards annual limit of 1 mSv for public exposure, up to 80 km from the reactor.

Detection of Prompt Gamma Radiation Near the LVR-15 Research Reactor

2019 ◽  
Vol 186 (2-3) ◽  
pp. 244-248
Author(s):  
Ladislav Viererbl ◽  
Vít Klupák ◽  
Antonín Kolros ◽  
Hana Assmann Vratislavská ◽  
Zdena Lahodová
Keyword(s):  
Gamma Radiation ◽  
Energy Range ◽  
Research Reactor ◽  
Reactor Core ◽  
Pulse Height ◽  
Wide Energy Range ◽  
Prompt Gamma ◽  
Reactor Building ◽  
Wide Energy ◽  
Prompt Gamma Radiation

Abstract The paper describes a method of pulse height spectrum measurement in a wide energy range. The LVR‑15 research reactor building was chosen to demonstrate this method. Pulse height spectra were measured on the third floor of the reactor building. Two types of scintillation detectors, NaI (Tl) and a plastic scintillator, were used. The detectors were placed for about 25 m from the reactor core, thus, separated from the primary circuit water in the reactor pool, biological shielding, building wall and other constructional materials. Spectra were measured in a wide energy range from 30 keV to 1000 MeV, in which signals were recorded from natural and man-made radionuclides, prompt gamma radiation and cosmic radiation. Experimental data were collected both while the reactor was in operation and while it was out of operation. This study confirms that differences in these spectra can be detected remotely over relatively large distances from the reactor core by adequately simple detection means.

scholarly journals RELAP5 SIMULATION FOR SEVERE ACCIDENT ANALYSIS OF RSG-GAS REACTOR

2018 ◽  
Vol 20 (1) ◽  
pp. 23 ◽  
Author(s):  
Andi Sofrany Ekariansyah ◽  
Endiah Puji Hastuti ◽  
Sudarmono Sudarmono
Keyword(s):  
Fuel Element ◽  
Secondary Flow ◽  
Research Reactor ◽  
Natural Circulation ◽  
Safety Analysis ◽  
Reactor Power ◽  
Severe Accident ◽  
Accident Analysis ◽  
Plate Temperature ◽  
The One

The research reactor in the world is to be known safer than power reactor due to its simpler design related to the core and operational chararacteristics. Nevertheless, potential hazards of research reactor to the public and the environment can not be ignored due to several special features. Therefore the level of safety must be clearly demonstrated in the safety analysis report (SAR) using safety analysis, which is performed with various approaches and methods supported by computational tools. The purpose of this research is to simulate several accidents in the Indonesia RSG-GAS reactor, which may lead to the fuel damage, to complement the severe accident analysis results that already described in the SAR. The simulation were performed using the thermal hydraulic code of RELAP5/SCDAP/Mod3.4 which has the capability to model the plate-type of RSG-GAS fuel elements. Three events were simulated, which are loss of primary and secondary flow without reactor trip, blockage of core subchannels without reactor trip during full power, and loss of primary and secondary flow followed by reactor trip and blockage of core subchannel. The first event will harm the fuel plate cladding as showed by its melting temperature of 590 °C. The blockage of one or more subchannels in the one fuel element results in different consequences to the fuel plates, in which at least two blocked subchannels will damage one fuel plate, even more the blockage of one fuel element. The combination of loss of primary and secondary flow followed by reactor trip and blockage of one fuel element has provided an increase of fuel plate temperature below its melting point meaning that the established natural circulation and the relative low reactor power is sufficient to cool the fuel element.Keywords: loss of flow, blockage, fuel plate, RSG-GAS, RELAP5 SIMULASI RELAP5 UNTUK ANALISIS KECELAKAAN PARAH PADA REAKTOR RSG-GAS. Reaktor riset di dunia diketahui lebih aman dari pada reaktor daya karena desainnya yang lebih sederhana pada teras dan karakteristika operasinya. Namun demikian, potensi bahaya reaktor riset terhadap publik dan lingkungan tidak bisa diabaikan karena beberapa fitur tertentu. Oleh karena itu, level keselamatan reaktor riset harus jelas ditunjukkan dalam Laporan Analisis Keselamatan (LAK) dalam bentuk analisis keselamatan yang dilakukan dengan berbagai macam pendekatan dan metode dan didukung dengan alat komputasi. Tujuan penelitian ini adalah untuk mensimulasikan beberapa kecelakaan parah pada reaktor RSG-GAS yang dapat menyebabkan kerusakan bahan bakar untuk memperkuat hasil analisis kecelakaan parah yang sudah ada dalam LAK. Simulation dilakukan dengan program perhitungan RELAP5/SCDAP/Mod3.4 yang memiliki kemampuan untuk memodelkan elemen bahan bakar tipe pelat di RSG-GAS. Tiga kejadian telah disimulasikan yaitu hilangnya aliran primer dan sekunder dengan kegagalan reaktor untuk dipadamkan, tersumbatnya beberapa kanal pendingin bahan bakar pada daya penuh, dan hilangnya aliran primer dan sekunder yang diikuti dengan tersumbatnya beberapa kanal pendingin bahan bakar setelah reaktor padam. Kejadian pertama akan membahayakan pelat bahan bakar dengan naiknya temperatur kelongsong hingga titik lelehnya yaitu 590 °C. Tersumbatnya satu atau beberapa kanal pada satu elemen bahan bakar menyebabkan konsekuensi yang berbeda pada pelat bahan bakar, dimana paling sedikit tersumbatnya 2 kanal akan merusak satu pelat bahan bakar, apalagi tersumbatnya satu elemen bahan bakar. Kombinasi antara hilangnya aliran pendingin primer dan sekunder yang diikuti dengan tersumbatnya satu kanal bahan bakar setelah reaktor dipadamkan menyebabkan naiknya temperatur kelongsong di bawah titik lelehnya yang berarti sirkulasi alam yang terbentuk dan daya yang terus turun cukup untuk mendinginkan elemen bahan bakar.Kata kunci: kehilangan aliran, penyumbatan, pelat bahan bakar, RSG-GAS, RELAP5

scholarly journals GASEOUS RELEASES EVALUATION AND SAFETY PERFORMANCE IMPROVEMENT OF KARTINI RESEARCH REACTOR VENTILATION SYSTEM

2012 ◽  
Vol 14 (2) ◽  
Author(s):  
Syarip Syarip ◽  
Widyatmaka Susyanta ◽  
Hadi Kusuma
Keyword(s):  
Fission Product ◽  
Radiation Exposure ◽  
Performance Improvement ◽  
Research Reactor ◽  
Source Term ◽  
Ventilation System ◽  
Ground Level ◽  
Safety Performance ◽  
Ground Level Concentration ◽  
Reactor Building

GASEOUS RELEASES EVALUATION AND SAFETY PERFORMANCE IMPROVEMENT OF KARTINI RESEARCH REACTOR VENTILATION SYSTEM. The safety performance of Kartini research reactor related to the gaseous releases to the environment has been evaluated. The research covers an evaluation and improvement on the ventilation system and analysis of gas releases dissipating from the reactor building. The method used is calculation of reactor source term and direct measurement of gas release from the reactor stack. The source term analysis showed that the fission product accumulated in the reactor core at the start of operation was 4.838 ´ 106 Ci, after of 5 hours operation it became 3.614 ´ 108 Ci, and after 24 hours decay, the fission product became 4.727 ´ 106 Ci. The N16 activity inside the reactor building is 4.1 ´ 10-10 μCi/cm3 and the Ar41 escaping to the atmosphere is 5.7 ´ 10-12 mCi/cm3, which is lower than limit value for radiation worker of 2 ´ 10-6 μCi/cm3. A sample case by using March 2009 data, the value of ground level concentration on variable distance x = 100 m to 5.000 m, was 9.726 ´ 10-19 rad/m3, rise up to 6.303 ´ 10-14 rad/m3 and tends to decrease to 1.598 ´ 10-15 rad/m3 at distance 5,000 m. Whiles the direct observation on the upper reactor stack show that the radiation exposure is 2.33 ´ 10-9 rad/s, exit velocity of gas from stack is 8 m/s, absolute temperature effluent of gas is 26.2 oC, and outlet diameter of stack, d = 1 m and actual stack height 31.75 m. Based on safety limit criteria from national regulation (BAPAETEN), the values of radiation exposure, ground level concentration combined with atmosphere stability and demography factor was very safe for the actual condition of Kartini reactor site. Keywords: safety performance, Kartini reactor, source term, ventilation system.

scholarly journals The EC MUSA Project on Management and Uncertainty of Severe Accidents: Main Pillars and Status

Energies ◽  
2021 ◽  
Vol 14 (15) ◽  
pp. 4473
Author(s):  
Luis Enrique Herranz ◽  
Sara Beck ◽  
Victor Hugo Sánchez-Espinoza ◽  
Fulvio Mascari ◽  
Stephan Brumm ◽  
...  
Keyword(s):  
Source Term ◽  
Initial Boundary ◽  
Severe Accident ◽  
First Year ◽  
Model Parameters ◽  
Spent Fuel ◽  
Horizon 2020 ◽  
Severe Accidents ◽  
Current State ◽  
Accident Scenarios

In the current state of maturity of severe accident codes, the time has come to foster the systematic application of Best Estimate Plus Uncertainties (BEPU) in this domain. The overall objective of the HORIZON-2020 project on “Management and Uncertainties of Severe Accidents (MUSA)” is to quantify the uncertainties of severe accident codes (e.g., ASTEC, MAAP, MELCOR, and AC2) when modeling reactor and spent fuel pools accident scenarios of Gen II and Gen III reactor designs for the prediction of the radiological source term. To do so, different Uncertainty Quantification (UQ) methodologies are to be used for the uncertainty and sensitivity analysis. Innovative AM measures will be considered in performing these UQ analyses, in addition to initial/boundary conditions and model parameters, to assess their impact on the source term prediction. This paper synthesizes the major pillars and the overall structure of the MUSA project, as well as the expectations and the progress made over the first year and a half of operation.

Mitigation of PWR In-Containment Iodine Source Term Under LOCA

2012 ◽  
Author(s):  
Khurram Mehboob ◽  
Xinrong Cao ◽  
Majid Ali ◽  
Rehan Khan
Keyword(s):  
Fission Product ◽  
Source Term ◽  
Retention Factor ◽  
Severe Accident ◽  
Modeling And Simulations ◽  
Surrounding Environment ◽  
Reactor Building ◽  
Accident Conditions ◽  
Safety Features ◽  
Containment Integrity

Since, containment integrity is the main issue under accidental conditions. Radiological consequences of LWR under accident have the grievous impact on the reactor building and its surrounding environment. Iodine is one of the most hazardous fission product releases in the serious accidents. So in this paper, the iodine source term has been evaluated for two-loop PWR under severe accident initiated due to LOCA. The TMI-2 reactor is considered as the reference reactor. The modeling and simulations are carried out by developing a MATLAB base program that uses the post-accident conditions and core inventory as input. The containment response, in order to mitigate the environmental and in-containment iodine source term is studied in normal, emergency, and isolation states of containment. The In-containment iodine source term is calculated with, and without the operation of engineering safety features (ESFs). The mitigation is determined by the activation of ESF. The environmental iodine source term is calculated as the function of containment response. The iodine dependency on the containment retention factor is also studied in all said states of containment. Results indicate the weak sensitivity of Iodine with activation of ESF towards exhaust rate values, under ESFs Operation.

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