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.

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.

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.

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 CALCULATION OF RADIOACTIVE SOURCE TERM RELEASE FROM FLEXBLUE NPP

2021 ◽  
Vol 23 (2) ◽  
pp. 63
Author(s):  
Muhammad Budi Setiawan ◽  
Pande Made Udiyani
Keyword(s):  
Nuclear Power ◽  
Noble Metals ◽  
Nuclear Reactor ◽  
Source Term ◽  
Reactor Core ◽  
Fission Products ◽  
Operating Conditions ◽  
Normal Operating ◽  
Small Modular Reactor ◽  
Mechanistic Approach

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).

Evaluation of Radiation Effects on Residents Living Around the NSRR Under External Hazards

2020 ◽  
Vol 6 (2) ◽  
Author(s):  
Yuiko Motome ◽  
Yoshiya Akiyama ◽  
Hiroyuki Murao
Keyword(s):  
Radiation Effects ◽  
Research Reactor ◽  
Reactor Core ◽  
Fission Products ◽  
Internal Exposure ◽  
Spent Fuel ◽  
Loss Of Function ◽  
Safety Research ◽  
Spent Fuel Storage ◽  
Fuel Storage

Abstract The nuclear safety research reactor (NSRR) is a research reactor of training research isotopes general atomics—annular core pulse reactor (TRIGA-ACPR) type, located in the Nuclear Science Research Institute (NSRI). The NSRR facility has been utilized for fuel irradiation experiments to study the behaviors of nuclear fuels under reactivity-initiated accident (RIA) conditions. Under the new regulation standards, which was established after the Fukushima Daiichi accident, research reactors are regulated based on the risk of the facilities. The graded approach is introduced in the regulation. To apply the graded approach, the radiation effects on residents living around the NSRR under the external hazards were evaluated, and the level of the risk of the NSRR facility was investigated. This paper summarizes the result of the evaluation in the case where the safety functions are lost due to a tornado, an earthquake followed by a tsunami. There is fuel in the reactor core, fresh fuel storage, and spent fuel storage. As the effects from reactor core, we evaluate the external exposure to radiation and exposure from the release of fission products assuming that loss of function to shut down the reactor, break of cladding tubes, loss of reactor pool water, and collapse of the reactor building. As the effects from fresh fuel storage, we evaluate the internal exposure assuming that the fresh fuel particles released into the air because of breaking into pieces. In addition, we evaluate the critical safety assuming that the critical safety shapes of the fresh fuel storage are lost. As the effects from spent fuel storage, we evaluate the critical safety assuming that the critical safety shapes of the spent fuel storage are lost. All in all, the risk is confirmed to be relatively low, since the effective dose on the residents is found to be below 5 mSv per event due to the loss of the safety functions caused by the tornado, earthquake, and the accompanying tsunami.

Best Estimate Analysis of PHEBUS FPT1 Test Using RELAP/SCDAPSIM Code

2012 ◽  
Author(s):  
Tadas Kaliatka ◽  
Eugenijus Ušpuras ◽  
Virginijus Vileiniškis
Keyword(s):  
Reactor Core ◽  
Fission Products ◽  
Severe Accident ◽  
Cooperative Research ◽  
Best Estimate ◽  
Fuel Bundle ◽  
Calculation Results ◽  
Estimate Method ◽  
Analysis Computer ◽  
Accident Conditions

The PHEBUS-FP program is an outstanding example of an international cooperative research program that is yielding valuable data for validating severe accident analysis computer codes. The main objective of the PHEBUS FPT1 experiment was to study the processes in the overheated reactor core, release of fission products and their subsequent transport and deposition under conditions representative of a severe accident of a Pressurised Water Reactor. The FPT1 test could be divided in the bundle degradation, aerosol, washing and chemistry phases. The objective of this article is the best estimate analysis of the bundle degradation phase. GRS (Germany) best estimate method with the statistic tool SUSA used for uncertainty and sensitivity analysis of calculation results and RELAP/SCDAPSIM code, designed to predict the behaviour of reactor systems during severe accident conditions, was used for the simulation of this test. The RELAP/SCDAPSIM calculation results were compared with the experimental measurements and calculations results, received by employing ICARE module of ASTEC V2 code. The performed analysis demonstrated, that the best estimate method, employing RELAP/SCDAPSIM and SUSA codes, is capable to model main severe accidents phenomena in the fuel bundle during the overheating and melting of reactor core.

IRSN R&D Actions on FP Behaviour for RCS, Containment and FCVS in Severe Accident Conditions

2016 ◽  
Author(s):  
Laurent Cantrel ◽  
Thierry Albiol ◽  
Loïc Bosland ◽  
Juliette Colombani ◽  
Frédéric Cousin ◽  
...  
Keyword(s):  
Porous Materials ◽  
Nuclear Power ◽  
Source Term ◽  
Cooling System ◽  
Fission Products ◽  
Simulation Software ◽  
Severe Accident ◽  
Status Report ◽  
Diagnosis And Prognosis ◽  
Accident Conditions

This paper deals with near past, ongoing and planned R&D works on fission products (FPs) behaviour in Reactor Cooling System (RCS), containment building and in Filtered Containment Venting Systems (FCVS) for severe accident (SA) conditions. For the last topic, in link with the Fukushima post-accident management and possible improvement of mitigation actions for such SA, the FCVS topic is again on the agenda (see Status Report on Filtered Containment Venting, OECD/NEA/CSNI, Report NEA/CSNI/R(2014)7, 2014.) with a large interest at the international scale. All the researches are collaborative works; the overall objective is to develop confident models to be implemented in ASTEC SA simulation software. After being initiated in the International Source Term Program (ISTP), researches devoted to the understanding of iodine transport through the RCS are still ongoing in the frame of a bilateral agreement between IRSN and EDF with promising results. In 2017, a synthesis report of the last 10 years of researches, which have combined experimental and fundamental works based on the use of theoretical chemistry tools, will be issued. For containment, the last advances are linked to the Source Term Evaluation and Mitigation (STEM) OECD/NEA project operated by IRSN. The objective of the STEM project was to improve the evaluation of Source Term (ST) for a SA on a nuclear power plant and to reduce uncertainties on specific phenomena dealing with the chemistry of two major fission products: iodine and ruthenium. More precisely, the STEM project provided additional knowledge and improvements for calculation tools in order to allow a more robust diagnosis and prognosis of radioactive releases in a SA. STEM data will be completed by a follow-up, called STEM2, to further the knowledge concerning some remaining issues and be closer to reactor conditions. Two additional programmes deal with FCVS issues: the MItigation of outside Releases in the Environment (MIRE) (2013–2019) French National Research Agency (NRA) programme and the Passive and Active Systems on Severe Accident source term Mitigation (PASSAM) (2013–2016) European project. For FCVS works, the efficiencies for trapping iodine with various FCVS, covering scrubbers and dry filters, are examined to get a clear view of their abilities in SA conditions. Another part, performed in collaboration with French universities (Lorraine and Lille 1), is focused on the enhancement of the performance of these filters with specific porous materials able to trap volatile iodine. For that, influence of zeolites materials parameters (nature of the counter-ions, structure, Si/Al ratio …) will be tested. New kind of porous materials constituted by Metal organic Frameworks (MOF) will also be looked at because they can constitute a promising way of trapping.

scholarly journals ESTIMATION OF THE RADIOACTIVE SOURCE TERM FROM RDE ACCIDENT POSTULATION

2019 ◽  
Vol 21 (3) ◽  
pp. 113
Author(s):  
Pande Made Udiyani ◽  
Ihda Husnayani ◽  
Mohamad Budi Setiawan ◽  
Sri Kuntjoro ◽  
Hery Adrial ◽  
...  
Keyword(s):  
Source Term ◽  
Noble Gas ◽  
Thermal Power ◽  
Reactor Core ◽  
Radioactive Source ◽  
Water Ingress ◽  
Calculation Results ◽  
Source Term Estimation ◽  
The Difference ◽  
Radioactive Release

The design process of Experimental Power Reactor (Reaktor Daya Eksperimental/RDE) has been carried out by BATAN for the last five years, adopting HTGR-type reactor with thermal power of 10 MW. RDE is designed with the reference of similar reactor, namely HTR-10. During this process, source term estimation is required to prove the safety of RDE design, as well as to fulfill the concept of As Low As Reasonably Achievable (ALARA) in radiation protection. The source term is affected by the magnitude of the radioactive substances released from the reactor core due to an accident. Conservative accident postulations on the RDE are water ingress and depressurization accidents. Based on these postulations, source term estimation was performed. It follows the mechanistic source term flow, with conservative assumptions for the radioactive release of fuel into the coolant, reactor building, and finally discharged into the environment. Assumptions for the calculation are taken from conservative removable parameters.The result of source term calculation due to the water ingress accident for Xe-133 noble gas is 8.97E+12 Bq, Cs-137 is 3.59E+07 Bq, and I-131 is 4.34E+10 Bq. As for depressurization accident, the source term activity for Xe-133 is 3.90E+13Bq, Cs-137 is 1.56E+07 Bq, and I-131 is 1.89E+10Bq. The source term calculation results obtained in this work shows a higher number compared to the HTR-10 source term used as a reference. The difference is possibly due to the differences in reactor inventory calculations and the more conservative assumptions for source term calculation.Keywords: RDE, HTGR, Radioactive, Source term, accident

Evaluation of Iodine Source Term in PWR under Loss of Coolant Accident

2012 ◽  
Vol 482-484 ◽  
pp. 1115-1119 ◽  
Author(s):  
Khurram Mehboob ◽  
Xin Rong Cao
Keyword(s):  
Nuclear Power ◽  
Source Term ◽  
Fission Products ◽  
Retention Factor ◽  
Severe Accident ◽  
Loss Of Coolant Accident ◽  
Loss Of Coolant ◽  
Term Evaluation ◽  
Computer Based ◽  
Instantaneous Release

During the severe accident in nuclear power plant (NPP), large amounts of fission products are released with accident progression, including In-vessel and Ex-vessel release. Thus, the Source term evaluation is essential for the probability risk assessment (PRA) and is still imperative for the licensing and operation of NPPs. Iodine is one of the most reactive fission products emitting in a large amount to containment and have a severe impact on health and sounding environment. Therefore, the iodine source term has been evaluated for 1000MW Reactor, by considering the TMI-2 as the reference reactor. The modeling and simulation of released radioactivity have been carried out by developing a MATLAB computer-based program. For post 1100 operation days, with the instantaneous release of radioactivity to the containment has been studied under LOCA. The dependency of radioiodine on ventilation exhaust rates has been studied in normal, emergency and isolation mode of containment. Moreover, the containment retention factor is also evaluated in said states of containment.

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