Juan Carlos Mora

Spanish experience on modeling of environmental radioactive contamination due to Fukushima Daiichi NPP accident using JRODOS.

Since the Chernobyl Nuclear Power Plant accident, decision support systems (DSS) for supporting response of the decision makers in emergencies have been developed and refined. Data available from real accidents are used to validate these systems, thus demonstrating their real capabilities and finally to improve them. This article presents the findings of the simulation exercises using JRODOS DSS performed in Spain after the first days of the accident in the Fukushima Daiichi Nuclear Power Plant. The investigation was carried out in two phases. The first phase is considered the early phase of the accident when few details of the real emissions are known (operational modeling). The second phase demonstrates how real measurements could be used (reconstructive modeling) to improve model predictions. Only major releases to the atmosphere, occurring during the first two weeks, were taken into account. Validation of the model was performed by direct comparison of the modeled results with real measurements.

Assessment for the management of NORM wastes in conventional hazardous and nonhazardous waste landfills.

Naturally Occurring Radioactive Materials (NORM) wastes are generated in huge quantities in several industries and their management has been carried out under considerations of industrial non-radioactive wastes, before the concern on the radioactivity content was included in the legislation. Therefore these wastes were conditioned using conventional methods and the waste disposals were designed to isolate toxic elements from the environment for long periods of time. Spanish regulation for these conventional toxic waste disposals includes conditions that assure adequate isolation to minimize the impact of the wastes to the environment in present and future conditions. After 1996 the radiological impact of the management of NORM wastes is considered and all the aspects related with natural radiations and the radiological control regarding the management of residues from NORM industries were developed in the new regulation. One option to be assessed is the disposal of NORM wastes in hazardous and non-hazardous waste disposals, as was done before this new regulation. This work analyses the management of NORM wastes in these landfills to derive the masses that can be disposed without considerable radiological impact. Generic dose assessments were carried out under highly conservative hypothesis and a discussion on the uncertainty and variability sources was included to provide consistency to the calculations.

Investigation of induced radioactivity in the linac-adone accelerator complex for the management of the decommissioned material

The LINAC-ADONE accelerator complex of the INFN-LNF Frascati National Laboratories, operating for 27 y prior to the commissioning of DA&PHgr;NE, was dismantled in 1993. The scraps resulting from the decommissioning of LINAC-ADONE have been temporarily stored in the same Frascati laboratory, waiting for definitive disposal. Relying on recommendations of the IAEA, European Commission and Italian committees, an experimental characterization study of the LNF repository was performed. The main objective was a classification of the scraps on the basis of internationally recognized “clearance levels,” which are 0.1 Bq g−1 for the isotopes of interest for this work. Secondly, a measurement of the materials suspected to be above 0.1 Bq g−1 was planned. Activation isotopes were expected from the aluminum, copper, steel, and iron of the LINAC and the ADONE ring sections. For screening purposes, the repository area has been divided into zones, where in-situ measurements with a portable HP-Ge detector have been performed. In addition, small samples have been cut from a representative number of pieces, and accurate laboratory measurements have been made with a low background HP-Ge spectrometer. The experimental results are in good agreement with other studies and show that a large part of the material is below the mentioned specific activity level.

Assessment of Dose and Risk for Public from Potential Exposure Using WinMACCS and JRODOS Codes

Potential exposure is prospectively considered exposure that is not expected to be delivered with certainty but that may result from an anticipated operational occurrence or accidents because of equipment failures, operating errors or external initiators. Potential exposure situations are events of probabilistic nature and the possible radiological impact as a comprehensive view shall be considered. In this paper a Probabilistic Safety Assessment (PSA) Level-3 for potential exposure was fulfilled using computer code MACCS (MELCOR Accident Consequence Code Systems). Additionally the deterministic modeling of consequence analysis for the critical meteorological conditions was fulfilled by the JRODOS decision support system (Real-time On-line Decision Support system for off-site emergency management in Europe). The framework for doses and risk assessment from potential exposure of accident releases are presented. Two approaches of safety acceptability analysis are demonstrated: the estimated doses dependent on distances of a release and other one is risk estimation. Both approaches are complementary, although the risk approach takes into account more aspects. So, the usage both of them can be considered an advantage. Comparison of the consequences with the risk curve acceptability criteria is shown.

Methodology Used in the Radiological Assessment of a Coal‐Fired Power Plant

A radiological assessment of the workers and the public potentially affected by the operation of the Teruel Coal‐fired Power Plant (the UPT Teruel), was performed under realistic assumptions. This assessment is part of a wider study to characterize the potential radiological impact of Naturally Occurring Radioactive Materials (NORM), in which our team, integrated by University of Extremadura and CIEMAT, is carrying out the study on coal‐fired power plants sponsored by the Spanish Nuclear Safety Council (CSN). The study comprises the four biggest coal‐fired power plants in Spain. Taking into account the working conditions and the plant specifications, six groups of workers were defined, established considering the 17 working tasks that could be of any importance for this assessment. For the public, considering that the area is barely inhabited, two different recreational scenarios were defined. Therefore, in‐plant and outside measurements, needed for the assessment of each scenario, were carried out. Where...

Probabilistic Prognosis of Environmental Radioactivity Concentrations due to Radioisotopes Discharged to Water Bodies from Nuclear Power Plants

Due to their very low values, the complexity of comparing the contribution of nuclear power plants (NPPs) to environmental radioactivity with modeled values is recognized. In order to compare probabilistic prognosis of radioactivity concentrations with environmental measurement values, an exercise was performed using public data of radioactive routine discharges from three representative Spanish nuclear power plants. Specifically, data on liquid discharges from three Spanish NPPs: Almaraz, Vandellós II, and Ascó to three different aquatic bodies (river, lake, and coast) were used. Results modelled using generic conservative models together with Monte Carlo techniques used for uncertainties propagation were compared with values of radioactivity concentrations in the environment measured in the surroundings of these NPPs. Probability distribution functions were inferred for the source term, used as an input to the model to estimate the radioactivity concentrations in the environment due to discharges to the water bodies. Radioactivity concentrations measured in bottom sediments were used in the exercise due to their accumulation properties. Of all the radioisotopes measured in the environmental monitoring programs around the NPPs, only Cs-137, Sr-90, and Co-60 had positive values greater than their respective detection limits. Of those, Sr-90 and Cs-137 are easily measured in the environment, but significant contribution from the radioactive fall-out due to nuclear explosions in the atmosphere exists, and therefore their values cannot be attributed to the NPPs. On the contrary, Co-60 is especially useful as an indicator of the radioactive discharges from NPPs because its presence in the environment can solely be attributed to the impact of the closer nuclear facilities. All the modelled values for Co-60 showed a reasonable correspondence with measured environmental data in all cases, being conservative in two of them. The more conservative predictions obtained with the models were the activity concentrations in the sediments of a lake (Almaraz) where, on average, values two times higher were obtained. For the case of rivers (Ascó), calculated results were adequately conservative—up to 3.4 times on average. However, the results for coasts (Vandellos II) were in the same range as the environmental measurements, obtaining predictions that are only—at maximum—1.1 times higher than measured values. Only for this specific case of coasts could it be established that the models are not conservative enough, although the results, on average, are relatively close to the real values.

Probabilistic risk assessment from potential exposures to the public applied for innovative nuclear installations

Potential exposures are those that may occur as a result of unanticipated operational performance or accidents. Potential exposure situations are probabilistic in nature because they depend on uncertain events such as equipment failure, operator errors or external initiators beyond the control of the operator. Consequently, there may exist a range of possible radiological impacts that need to be considered. In this paper a Level 3 Probabilistic Safety Assessment (PSA) for a hypothetical scenario relevant to Innovative Nuclear Energy Systems (INS) was conducted using computer code MACCS (MELCOR Accident Consequence Code Systems). The acceptability of an INS was analyzed taking into account the general requirement that relocation or evacuation measures must not be necessary beyond the site boundary. In addition, deterministic modeling of the accident consequences for the critical meteorological conditions was carried out using the JRODOS decision support system (Real-time On-line Decision Support system for off-site emergency management in Europe). The approach used for dose and risk assessment from potential exposure of accidental releases and their comparison with acceptance criteria are presented. The methodology described can be used as input to the licensing procedure and engineering design considerations to help satisfy relevant health and environmental impact criteria for fission or fusion nuclear installations.