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In CANDU type reactors there are two situations for irradiated fuel bundles to remain accidentally in the air. One is represented by End Fitting Failure accident when one or more fuel bundles are ejected from the channels during the fuel handling process. The second one is determined by the uncovering of fuel bundles in the Spent Fuel Pool (SFP) due to a partial or total loss of cooling followed by a progressive evaporation of the water. In both cases there is a potential for the release of radio-nuclides that is dependent on the irradiation history and the duration between unloading of the bundle from the core and the entering into the atmosphere. The paper is devoted to investigate the release in these cases. The inventory at the moment of the accidentally entering into the atmosphere is calculated by ORIGEN code. The source term into the containment, and respectively in the SFP building, is obtained by treating the fuel degradation and the release from the fuel elements by ICARE module of ASTEC code. The most important influence factors on the source term formation (such as the velocity of the ventilation, the residence time) are identified and discussed. In a realistic CANDU SFP loss of cooling accident the potential of release is reduced due to relative fuel low burnup, long residence time, and specific configuration of the storage. The uncovering of the first plane of fuel bundles needs around 15 days for the heat-up and evaporation of water. In such case, a small convection ( v  ≥ 1.E−05 m/s) is enough to keep the temperature of the claddings under the failure thresholds. The paper identifies and investigates some other cases (with violation of the operational procedures) with a relevant release of fission products. The results are presented for the Caesium element. [ABSTRACT FROM AUTHOR]

Ghana is in the process of converting its fuel from Highly Enriched Uranium (HEU) to Lowly Enriched Uranium (LEU). Radioactive fission and activation products generated in the irradiated nuclear fuel are hazardous to personnel, environment and the public. Investigation into the levels of radiation dose will help in a safe and efficient core conversion process. Two computer codes used were ORIGEN-S; for computing changes in the isotopic concentrations during neutron irradiation and radioactive decay as well as to determine the source term and MCNP6; which used the source term estimated by ORIGEN-S code to calculate the dose rate. The criticality of the core at different heights above the bottom of the core was also obtained. Most of the radionuclides present after the core depletion contributed to the source term of 1.767 × 10 13 ± 0.0008 photons/sec which was observed after thirty days of the cooling period. The dose rates ranged between 3.51 × 10 −25 ± 0.0003 mGy/h and 4.27 × 10 4 ± 0.0006 mGy/h at different positions above the reactor core, the control room (wall, door and window) and the rabbit room. The criticality (k eff ) also decreased from 0.99442 ± 0.00006 to 0.01238 ± 0.00002 as the core moved from the bottom of the reactor vessel to the top. [ABSTRACT FROM AUTHOR]

This paper presents and discusses the evaluation of core inventory, the source term (ST) and doses results during and after a hypothetical accident at the NUR nuclear research reactor (RR). In this study, a systematic approach has been adopted for the necessary calculations using specialized computer codes following the latest safety recommendations of the International Atomic Energy Agency (IAEA). The Oak Ridge Isotope Generation computer code version 2 (ORIGEN2) is used for core inventory calculations for 162 full-power post-operation continuous days. The accident sequences suppose the destruction of all fuel elements by a plane crash. The atmospheric dispersion modeling of radionuclides was modelled using health physics code version 3.01 (HOTSPOT3.01) in order to estimate the total effective doses equivalent (TEDE) in the air and on the ground, and enable the assessment of the contamination of individuals receiving doses in an area defined by a radius of 2 km downwind from the NUR reactor building vicinity. The results can also be used to define adequate protective measures for emergency situations. [ABSTRACT FROM AUTHOR]

Ruthenium is one of the most radiotoxic fission products which can be released from fuel as ruthenium oxides in an air ingress accident at a nuclear power plant. In this study it was found that the transport of the released ruthenium oxides through a reactor coolant system into the containment building is significantly affected by the atmospheric conditions. Airborne CsI increased the transport of gaseous ruthenium compared with that in a pure air atmosphere. The overall transport of ruthenium increased with temperature. In order to understand the behaviour of ruthenium in accident conditions, it is important to widen the experimental conditions from pure air/steam atmospheres to more realistic mixtures of prototypic gases and aerosols. [ABSTRACT FROM AUTHOR]

In this paper, a small PWR with no pressure vessel external cooling system (ex-vessels) is used as the research object. It is found that for the small low burnup marine reactor without ex-vessels, the more core releases at an early stage, the later core damage occurs. Considering in the case of large break loss of coolant and loss of power accident, and the ORIGEN program is coupled with the MELCOR program is used as a severe accident process simulation tool. The significant parameters have been took into account, include core melting time, core damage rate, and lower head failure time, etc. It could be found through calculation that, for a certain reactor, in case of the reactor core meltdown happened, different released amount of source term will results in a difference of core heat release in the pressure vessel. For the small reactors without ex-vessels, the amount difference of source term release can reach 20%, and the lower head failure time can be delayed by 70%. • Radioactivity release from small low burnup Marine reactors. • Three radioactive release models affect accident consequences. [ABSTRACT FROM AUTHOR]

Radioactive waste management is a strategic activity that is adopted in the nuclear field to safeguard the operating staff, the population and the environment from radiological risks. Studies that support the environmental assessment of a nuclear site offer insight into the understanding of the transport of radionuclides in environmental matrices (e.g. soil, groundwater, surface water, etc.). These studies, that involve the migration of a radiological source in the environment, are conducted to identify how to approach the safety assessment of current or future nuclear facilities. This work analyzes the source term as the key point in the modeling of the transport of radionuclides in groundwater and soil. A methodological approach, which focuses on the dynamic of the source term in space and time, was applied to model the transport of radionuclides. This approach can be used to plan reliable environmental monitoring networks. The analysis was performed at the nuclear site of Saluggia, Vercelli (Italy). [ABSTRACT FROM AUTHOR]

During a hypothetical severe accident in a nuclear reactor most of the radioactivity would be released from the damaged core and the broken primary circuit in the form of aerosols and chemical vapours and gases. However the effective radio-toxicity per unit release from the fuel depends both on the released mass and on the chemical form as well. The fission products release and transport in the circuit has been analysed for a Medium Break cold leg Loss Of Coolant Accident (MBLOCA) combined with the total loss of electric power supply (Station Black Out) for a reference PWR 1000 reactor using the severe accident integral code ASTECV1.3 rev2. A detailed description of the reactor model is given. Sensitivity analyses were performed in terms of break location and accumulator availability. The break location has an impact on the carrier gas temperature and in turn on the chemical speciation of the fission products. Also retention is influenced by the different flow pathways establishing in the circuit. The accumulator availability extends the duration of the core degradation process affecting the total production of hydrogen and the cumulative released mass of low, semi and highly volatile fission products. In agreement with the ASTEC equilibrium chemistry model, the Mo/Cs (Rb) molar ratio has the most influence on the fraction of iodine which can reach the containment in the gaseous or highly volatile form. Due to the surplus of Cs, the predicted dominant iodine species to the containment are CsI and its dimers. A 3–4% of iodine which reaches the break exists in gaseous (HI) and highly volatile metal-iodides (I 2 MoO 2 , SnI 2 , SnI 4 ) species. [ABSTRACT FROM AUTHOR]

Abstract: The possibility of an accident or component failure during mid-loop operation has been identified in probabilistic safety studies as a major contributor to core melt frequency and source term risk. The fission products release and transport to the containment has been analyzed during mid-loop operation of a reference PWR 1000 MWe reactor using the severe accident integral code ASTEC V2.0. The analyses have been performed considering the loss of residual heat removal (RHR) system at various times after reactor shutdown for the reactor vessel configuration with the removed upper head (open reactor). In this configuration, the possible air ingress can have an impact on safety such as accelerated oxidation and increased volatility of certain FPs (particularly iodine and ruthenium). Sensitivity calculations have been performed in terms of air ingress simulation with a different intensity. Besides equilibrium chemistry model, most of the calculations have also used a limited kinetics model. The study has shown that without air ingress the only predicted gaseous form of iodine is HI (≤7.4% of the total mass of iodine released from core) and no gaseous RuO4 is created. Sensitivity calculations have illustrated that the gross fraction of gaseous iodine (I2 + HOI + HI) has an increased trend with growth of air ingress intensity and with the duration of sequence evolution. In most oxidative atmosphere the gross iodine gaseous fraction could increase by a factor form of two to several times as compared to the corresponding case without air ingress (particularly due to I2 persistence). Creation of gaseous RuO4 is sensitive to carrier gas temperature; therefore a considerable fraction (≤3%) is predicted only in the sensitivity cases with the shortest time of loss of RHR after reactor scram. [Copyright &y& Elsevier]

Abstract: The 6th FWP SARNET project launched a set of studies to enhance understanding and predictability of relevant-risk scenarios where uncertainties related to aerosol phenomena were still significant: retention in complex structures, such as steam generator by-pass SGTR sequences or cracks in concrete walls of an over-pressurised containment, and primary circuit deposit remobilization, either as vapours (revaporisation) or aerosols (resuspension). This paper summarizes the major advances achieved. Progress has been made on aerosol scrubbing in complex structures. Models based on empirical data (ARISG) and improvements to previous codes (SPARC) have been proposed, respectively, for dry and wet aerosol retention, but, further development and validation remains, as was noted during the ARTIST international project and potential successors. New CFD models for particle-turbulence interactions have been developed based on random walk stochastic treatments and have shown promise in accurately describing particle deposition rates in complex geometries. Aerosol transport in containment concrete cracks is fairly well understood, with several models developed but validation was limited. Extension of such validation against prototypic data will be feasible through an ongoing joint experimental program in the CEA COLIMA facility under the 6th Framework PLINIUS platform. Primary deposit revaporisation has been experimentally demonstrated on samples from the Phebus-FP project. Data review has pinpointed variables affecting the process, particularly temperature. Available models have been satisfactorily used to interpret separate-effect tests, but performing integral experiments, where revaporisation is likely combined with other processes, still pose a difficult challenge. Further experimental data as well as modelling efforts seem to be necessary to get a full understanding. Resuspension, sometimes referred to as mechanical remobilization, has been recently addressed in SARNET and although a set of models were already available in the literature (i.e., Rock''n Roll model, CESAR, ECART), further work is needed to extend current capabilities to multi-layer deposits and to produce simplified, but sufficiently accurate, models. A major remaining uncertainty is the particle-to-particle/wall adhesion and its dependence on microscale roughness. Data from the previous EU STORM project have been retrieved and further experiments designed for code validation are being used to benchmark the models. [Copyright &y& Elsevier]

Assessment of the potential radiological impact and emergency zones for small Integral Pressurized Water Reactors (iPWR) is very important for their employment in the near term. In this work, an attempt is made to provide an analysis and estimation of the envelope of the potential impact from a severe accident in an iPWR. The source term is calculated from the solution of the lumped aerosol concentration equation in the containment, by using different approaches for estimating the various parameters. The first approach comprises a straightforward use of the methodology and parameters used for large reactors differing only in the lower power level of an iPWR. To account for the anticipated enhanced aerosol retention in the containment and the slower core damage progression in case of iPWRs, two additional approaches are then used. The values of the deposition velocities required are specified using relevant data from available detailed computational and experimental studies. Conservative, yet representative, accidental conditions in the containment are specified, based on insights offered by available analyses with severe accident codes available in the open literature and disclosed regulatory safety assessments. The radiological impact is assessed using a Gaussian plume model, whereas the exposure is assessed against the latest reference levels and dose criteria. Fetus exposure was found to determine the level of the radiological impact during the plume phase. The radioactive releases and the associated radiological impact may differ decisively depending on the approach used to estimate the source term. Consideration of the enhanced safety attributes of iPWRs (namely, enhanced radioactivity retention in the containment and slower accident progression) leads to a reduction of more than one order of magnitude, from about 75 TBq to 4 TBq and 6 TBq to 0.3 TBq , in terms of I131 and Cs137 releases to the atmosphere, respectively. Correspondingly, emergency distances are estimated to span from about 4 km - a level commensurate with the generic distances suggested in international standards for low power nuclear reactors - to a distance as low as 600 m if the favorable characteristics of iPWRs are considered, thus implying that no significant measures are required off-site. • Estimated release was reduced by ten times due to iPWRs design characteristics. • Releases of volatiles is completed very quickly after core damage. • Fetus dose remains below criteria off-site, if iPWRs characteristics are considered. • Ignoring the favorable characteristics of iPWRs leads to a 4-km emergency zone. • The slower core damage offers more time for an optimized response off-site. [ABSTRACT FROM AUTHOR]

Simultaneous estimation of bioluminescence source term and boundary conditions from in situ irradiance data is presented. Inverse analysis is performed by solving a nonlinear optimization problem, where the objective function is given by the square difference between experimental and computed data plus a regularization term. The forward problem is tackled with the LTSN method that numerically solves the radiative transfer equation. The experimental data are simulated with synthetic data corrupted with noise. [Copyright &y& Elsevier]

The decomposition of deposited CsI and iodine oxides (IOx) aerosols deposited on sand bed and on a metallic filter has been pursued under irradiation at 120 °C. On the sand, it was found that CsI aerosols are decomposed by the irradiation and the temperature effect into an inorganic specie assumed to be molecular iodine. The temperature effect might be due to surface effects of the sand grains as CsI aerosols are usually stable at this temperature. IOx aerosols are expected to be well retained by the sand filter (DF > 100) as long as the sand height is high enough (>60 cm). IOx are also decomposed by the temperature and irradiation. The humidity is also expected to have a significant influence on their decomposition. On the metallic filter, despite both species are expected to be significantly decomposed, very few gaseous iodine release from the filter was observed. The gaseous iodine species produced by the aerosols decomposition is assumed to be quickly adsorbed on the steel wires even if a slow desorption is also expected. • Thermal and radiolytic decomposition of CsI and IOx aerosols deposited on sand and metallic filter was studied and quantified. • Decontamination factors (DF > 100) were determined for small aerosols like IOx (MMAD ≈ 0.16 μm) deposited on a sand filter whose thickness is over 60 cm. • On the sand filter, it is observed a significant CsI and IOx aerosols radiolytic decomposition leading to a release of gaseous inorganic iodine. A thermal effect is also observed and is assumed to occur at the surface of the sand grains surface. • On the metallic filter, CsI and IOx aerosols are expected to be significantly decomposed. but there is no significant release of gaseous iodine because of the large surface developed by the steel wires on which iodine is expected to be adsorbed. [ABSTRACT FROM AUTHOR]

Significant concerns after Fukushima disaster were raised on the potential of radioactive release from Spent Fuel Pool accidents. In case of CANDU type reactors a set of factors, such as the low discharge burnup of the spent fuel and the pool configuration with trays and racks placed on the bottom of the pool, contributes to the reducing of this potential. In this paper a peculiar state is analysed in relation with the refurbishment process, when a full core is added into the spent fuel pool, in a shorter time then the normal refuelling discharge. A prolonged loss of cooling is postulated in order to boil and evaporate the water. The claddings of the fuel bundles accidently entered in the air will be progressively heated and then degraded. Fission products will be released into the atmosphere of the spent fuel pool and, possibly, into the environment through the existing junctions. The paper investigates the source term formation at the level of the environment considering a prolonged accident without any interventions. The simulation is structured into three phases: de-watering of the pool, fuel degradation, transport of the radioisotopes to the environment. In order to estimate the maximum potential of the accident, three scenarios for the discharge of the core were considered, all of them requiring an enlarged cooling capacity in comparison with the normal refuelling, aiming a reduction of the duration of the refurbishment process to reduce the economic penalties. [ABSTRACT FROM AUTHOR]

This study aims at analysing the performance of the European Radiological Data Exchange Platform (EURDEP) at identifying, delineating, and tracking different radioactivity clouds moving across a territory. To this purpose, the atmospheric dispersion of 21 hypothetical STs, calculated by the Modular Accident Analysis Program (MAAP) code, has been simulated under the same meteorological scenario by the JAVA-based RODOS decision support system (JRODOS). EURDEP's performance is evaluated using spatial and temporal parameters. The present analysis shows a large impact of the total amount of activity and the timing of the release peak on EURDEP's performance. Monitoring stations are able to detect quite rapidly the release (first alarm), and presents more variation in the time needed to detect the peak, and to delineate and track the plume. A two-dimensional analysis is suggested to illustrate strengths and weaknesses of EURDEP to face different types of STs. These results highlight the need to link the understanding of ST characteristics with meteorological conditions to limit the environmental and health effects associated with radioactive emissions into the atmosphere. The present study aims at presenting a methodology to analyse the performance of EURDEP in other territories. • Evaluation the performance of the EURDEP system under different model predictions. • Atmospheric dispersion of 21 source terms combining JRODOS and MAAP. • Evaluation the network design in the early detection of radioactivity in the atmosphere. • Methodology to evaluate gamma dose rates networks. [ABSTRACT FROM AUTHOR]

This paper provides an overview of the current MSR design space and lists unique features of the various designs under consideration. Some general considerations for source terms calculation for Molten Salt Reactors (MSRs) are explained. Applicability and limitations of terminology currently defined for legacy light water reactor (LWR) systems are discussed in the view of MSRs and the need for updated terminology is discussed. Calculations carried out for the Molten Salt Reactor Experiment (MSRE) are discussed with a qualitative comparison to the designs presented. The nature of the fission products (FPs) and actinides for Low enriched uranium, thorium and fast U/Pu fuel cycles employed in representative molten salt reactor systems are discussed. Computational results are obtained from a code (Serpent 2) with online reprocessing. Divergence in source terms when fission product bubbling is demonstrated. The source release for each molten salt reactor during postulated accidents is also presented. Image 1 • Discussion on terminology for molten salt source terms and accident scenarios are presented. • Burnup calculations and assumptions for source term analysis is presented. • Three molten salt reactor designs are compared against each other with respect to source term release. [ABSTRACT FROM AUTHOR]