Low And Intermediate Level Radioactive Waste Research Articles

The 14C dose assessment model chain – 14C source term definition and uncertainty quantification

Abstract. Carbon-14 (14C) is an important, dose-determining contributing radionuclide in a deep geological repository for low and intermediate level radioactive waste (L/ILW). 14C is relevant for safety assessments as it can be present both as dissolved and gaseous species in the disposal facility and the host rock, and 14C-organic compounds migrate in the near field and the geosphere due to weak interaction with mineral surfaces in alkaline to near-neutral conditions. Thus, the chemical form of the 14C-bearing carbon species dictates the routes of 14C migration in the engineered barrier system of a deep geological repository and the surrounding host rock and therefore determines the long-term contribution of 14C to dose release from a repository for radioactive waste. Furthermore, 14C released from a deep repository would be subject to a certain accumulation in the biosphere, given the role of carbon as main building block of any life form on earth. The contribution will illustrate Nagra's recent progress within the 14C dose assessment model chain for a L/ILW deep geological repository to address the quantification of uncertainties. The systematic approach began by incorporating a realistic best-estimate plus uncertainty description of 14C activities within the Swiss model inventory of radioactive materials assembled specifically for the purpose of the general licence application. In this context state-of-the-art activation calculations for expected reactor and decommissioning wastes as well as extensive inquiries into origin and use of 14C within medicine, research, and industry played a key role. Uncertainties were screened for relevance and main contributing factors taken into account, differentiating between the characterisation methods of the relevant waste streams. Based on detailed knowledge of the expected activity and material inventory, release rates for the release pathways through congruent and instantaneous release processes as well as resulting chemical species could be defined with confidence. Key findings and characteristics of the 14C source term will be highlighted and discussed.

Research on the γ spectrum-unfolding method of low- and intermediate-level radioactive waste based on LaBr3(Ce) detector

To apply the LaBr3(Ce) detector with a higher full-energy peak detection efficiency and larger activity detection range to the measurement of low- and intermediate-level radioactive waste (LILW), we examined the problem of poor energy resolution. We analyzed the characteristics of key nuclide energy spectra and established an adaptive filter function to form a deconvolution energy spectrum analysis method for intelligent nuclide identification and quantitative calculation. Compared with the traditional deconvolution method, the average overlapping peak resolution was improved by 74.5%, achieving the qualitative analysis and quantitative calculation of all nuclides. By using the novel method to analyze the experimental spectrum of Co-60 and Cs-137, the results showed that the nuclide was accurately identified with a mean relative error of 0.85% for the peak area calculations. The analysis of the spectra of the compositions of the five LILW nuclides was obtained from Monte Carlo simulations using different methods. The results showed that, for the overlapping peaks with peak area ratios of 1:1:1:1:1, compared with the traditional deconvolution method, the peak area calculation error of this method was reduced from 73.09% to 2.12%. For the overlapping peaks with the peak area ratios of 5:1:2:3:1 and 5:1:20:10:1, the traditional spectrum-unfolding method could not resolve the weak peaks, and the maximum energy deviation was 8 keV. This method achieved accurate identification of all energies, and the error of the weak peak was 7.19%, and the error of the other peaks was less than 2.23%. The results of this study provided key technical support for the application of LaBr3(Ce) detectors to the nuclide measurement of LILW.

Power and participation in the field of radioactive waste disposal

ABSTRACT In the last decades of the 20th century, a ‘participatory turn’ could be detected in radioactive waste management (RWM). Participatory governance and public participation were put forward with the aim of democratizing policy and decision-making. As participatory processes are quite complex, an examination of how public participation is operationalized often offers no straightforward answers concerning the realizations of its democratizing intentions. This study contributes to understanding the ‘double bind’ of public participation. Moreover, we elaborate on how participation can lead to democratization and to the reinforcement of the status-quo. We draw on our experiences within a large-scale participatory project aimed at siting, developing, and building a repository for low and intermediate level radioactive waste (LILW) in Belgium. In our study, we combine a historical perspective of the already two decades-long participatory process with a micro-analysis of a specific participatory initiative.

Groundwater Flow Modeling of a Near-Surface Disposal Repository for Low- and Intermediate-Level Radioactive Waste in Southwest China

The influence and function of engineering facilities were increasingly concerned about the safety analysis of low- and intermediate-level radioactive waste (LILW). In terms of near-surface disposal, many artificial facilities were set down, such as drainage facilities, covering layers, and disposal units. To analyze the long-term impact of these artificial facilities on groundwater in the disposal site area, we built four time-continuing models by setting the time nodes of parameter or boundary changes using FEFLOW code, considering the possible aging and degradation state of these facilities. According to the models, the site area’s groundwater level situations for long-term safety assessment were predicted. The results showed the different regulating abilities of drainage facilities affected the groundwater level of the disposal site in different degradation states, which also reflected the necessity of artificial facilities simulation in groundwater modeling. In addition, the Monte Carlo method and surrogate model were adopted to analyze the influence of the uncertainty of model parameters on the output of groundwater flow models. This study could help in further understanding the groundwater flow modeling for long-term safety assessment of near-surface disposal engineering.

Emplacement optimization for the first-stage disposal facility of LILW complex disposal facility by employing simulated annealing algorithm

In order to optimize the emplacement of LILW (Low- and Intermediate-Level radioactive Waste) complex disposal facility, an emplacement optimization program POWER (Program for Optimization of Waste package Emplacement in Repository) based on Simulated Annealing (SA) algorithm is developed. The POWER program derives near-optimum emplacement schemes that minimize the objective function consisting of total dose, radiotoxicity flux, and radiotoxicity concentration obtained from GoldSim calculations for the post-closure safety assessment for LILW rock cavern disposal facility. An optimization problem of sixty nine drums with different degradation mechanism, source term, and drum type to be disposed of into the first-stage disposal facility is defined first and a series of optimization calculations is performed. It is verified that the POWER can generate the near-global optimized emplacement scheme from SA process.

Predicting degradation of organic molecules in cementitious media

Degradation of organic compounds is a key issue in the performance assessment programs for Low and Intermediate Level radioactive Waste (L-ILW) repositories. Despite significant progress in recent decades, parameterization and implementation of biotic degradation of these compounds in reactive transport models still remain a challenging task, mainly due to uncertainties linked to their complex chemical behaviour.In this work we have explored the application of thermodynamic models in order to stablish a predictive kinetic model for nineteen organic molecules of particular interest in the frame of Low and Intermediate Level Waste disposals. Firstly, from a compilation of experimental data taken from the literature, we observed a decreasing trend in the zero-order kinetic degradation rates as the number of carbon atoms increases. Even though this conclusion is based on thermodynamic and experimental data obtained at circumneutral pH conditions, it can be considered as indicators of most probable processes expected at foreseen alkaline storage conditions.Based on the model published by LaRowe and Van Cappellen (2011) for natural organic matter degradation, a set of organic molecules of interest were analysed and compared. This thermodynamic calculations allowed us to classify the molecules of interest in three different groups regarding their most probable behaviour: i) adipate, succinate, pimelate, benzoate, TPB− and DBP− been identified as persistent molecules as their degradation is thermodynamically inhibited and they represent potential radionuclides complexing agents; ii) oxalate, formate, gluconate, lactate, NTA−, propionate, acetate, ISA- and sorbate will be easily oxidized and their kinetic degradation will proceed while the substrate for bacterial growth remain available. A third group can be mentioned and comprises organic molecules having a partly inhibited degradation from the thermodynamic point of view (DBP− EDTA−, DTPA−). Different biodegradation pathways are mentioned in the literature for these molecules. Given the high complexity of the involved processes, more experimental studies are needed before concluding on their availability for microbial processes.

Safety assessment of second-phase disposal facility in Gyeongju low- and intermediate-level radioactive waste (LILW) repository using RESRAD-OFFSITE code

ABSTRACT A safety assessment was performed on the second-phase disposal facility of the Gyeongju low – and intermediate-level radioactive waste repository using the RESRAD-OFFSITE code. On the basis of the acceptance criteria of the repository, 13 radionuclides and gross alpha were selected as the radionuclides to be evaluated. In addition, the thermodynamic database of Japan Atomic Energy Agency and natural groundwater data that reflect the operational condition of the disposal facility were applied to derive the solubility by using the PHREEQC computational code. The distribution coefficient was derived for the component of disposal facility based on the sorption database of Japan Atomic Energy Agency. In addition to the solubility and distribution coefficient for various radionuclides, the data of the disposal facility and repository site were analyzed and applied to the RESRAD-OFFSITE code for the safety assessment modeling. The result of the safety assessment calculation revealed the exposure dose corresponding to 3.8% of the exposure dose limit (0.1 mSv/yr). Moreover, additional sensitivity analysis for the exposure dose was conducted by varying the geochemical parameters. Although the exposure dose changed as the geochemical parameters were varied, the result of the sensitivity analysis confirmed that the exposure dose limit was still satisfied.

A Study About Radionuclides Migration Behavior in Terms of Solubility at Gyeongju Low- and Intermediate-Level Radioactive Waste (LILW) Repository

A safety assessment of radioactive waste repositories is a mandatory requirement process because there are possible radiological hazards owing to radionuclide migration from radioactive waste to the biosphere. For a reliable safety assessment, it is important to establish a parameter database that reflects the site-specific characteristics of the disposal facility and repository site. From this perspective, solubility, a major geochemical parameter, has been chosen as an important parameter for modeling the migration behavior of radionuclides. The solubilities were derived for Am, Ni, Tc, and U, which were major radionuclides in this study, and on-site groundwater data reflecting the operational conditions of the Gyeongju low and intermediate level radioactive waste (LILW) repository were applied to reflect the site-specific characteristics. The radiation dose was derived by applying the solubility and radionuclide inventory data to the RESRAD-OFFSITE code, and sensitivity analysis of the dose according to the solubility variation was performed. As a result, owing to the low amount of radionuclide inventory, the dose variation was insignificant. The derived solubility can be used as the main input data for the safety assessment of the Gyeongju LILW repository in the future.

Chemical degradation of fly ash blended concrete with the seasonal variation of rainwater in a radioactive waste repository: A thermodynamic modeling approach

This study presents a long-term degradation behavior of fly ash blended concrete in a vault type low- and intermediate-level radioactive waste (LILW) repository by thermodynamic equilibrium calculations using PHREEQC combined with CEMDATA18 database. Since rainwater plays a predominant role as leachate for the concrete waste container in the repository, the effect of chemical properties of seasonal rainwater on the chemical degradation of concrete was analyzed. In addition, the impact of the blending with fly ash on the chemical degradation was evaluated through a gradual replacement of ordinary Portland cement (OPC) by fly ash. Regardless of the replacement level, autumn rainwater showed the greatest influence on the concrete degradation owing to the presence of inorganic carbon species, which induce the carbonation. The resistance of concrete to chemical degradation by rainwater was persistently reduced due to the decreasing formation of portlandite and calcium silicate hydrate (C-S-H) with increasing replacement level.

Geopolymerisation of gasified ion-exchange resins, mechanical properties and short-term leaching studies

A combination of gasification of low and intermediate level radioactive waste (LILW) and conditioning of the resulting product within a geopolymer matrix is a potential alternative for vitrification technologies as an immobilisation method. Geopolymer matrices have demonstrated good retention capability for radionuclides in many studies and the technology has been implemented at an industrial scale in Slovakia and Czech Republic. However, the practical waste loading has been limited by the mechanical properties of the encapsulated matrix. Even a small amount of ion exchange resin (IXR) decreases the strength of the matrix and cohesion of the matrix is lost when the fraction of resin exceeds 15-20%. In this study, the potential to combine gasification as thermal treatment and various inorganic binders as encapsulation matrices was evaluated. After gasification, the mechanical properties were not similarly sensitive to the encapsulation of IXR. Gasification enabled substantially higher loading of IXR into the sample. Also, gasification of the IXR decreased matrix apparent Caesium diffusion. Very low apparent diffusion coefficients of Cs were calculated when gasified resin was encapsulated in metakaolin matrix. Theoretically, the amount of Cs within the same volume of encapsulated material could be increased by 800 times following gasification and encapsulation within an alkali-activated metakaolin (MK) binder.

Spatiotemporal changes in hydrogeochemistry of coastal groundwater through the construction of underground disposal facility for low and intermediate level radioactive wastes in Korea

This study was objected to assess active changes in hydrochemistry of coastal groundwater through the construction of underground low- and intermediate-level radioactive waste (LILRW) disposal facility. Groundwater monitoring data from the construction stage of the facility beginning in 2010 to the operation stage in 2017 revealed seawater intrusion (SWI) into the inland area with significant spatiotemporal variation and subsequent changes in hydrochemical reactions. Groundwater was divided into three groups based on the temporal changes in the seawater mixing ratio. Group 1 showed an increasing trend with some delayed responses to SWI and an ion-exchange dominant process. Group 2 represented dynamic changes of hydrochemical reactions in time toward to a new equilibrium condition along SWI and refreshing. In this group, simultaneous increases in SO4 and Ca, depletion of δ34SSO4, and a slight decline in the SI of gypsum suggest that precipitation and re-dissolution of gypsum occurred during the transition of SWI. These changes are probably attributable to the balance change between seawater and fresh groundwater inflows, and subsequently to preferential flow paths developed in the fractured aquifer. Group 3 groundwater demonstrated a rapid and direct connection with seawater since the beginning of construction of the LILRW facility. In summary, SWI process and subsequent hydrogeochemical reactions in the coastal groundwater varies significantly in time and in space, thus continuous monitoring of groundwater chemistry is warranted for safety management of the underground LILRW facility.

Possibility of using SFR-SCC for the production of the LILW disposal container

The disposal container with low and intermediate level radioactive waste (LILW) was developed using self-consolidating concrete (SCC), which provides long term mechanical properties and extremely low permeability for liquids. In the paper, some results of additional investigations are presented, which complement the assessment of the behaviour of SCC during use. The main focus is the presentation and discussion of the results of SFR-SCC (steel fibre reinforced - self-consolidating concrete) tests in fresh and harden state. From these results it can be seen that there is a possibility of further improving the properties of the SCC and thus the container.

Radiotoxicity flux and concentration as complementary safety indicators for the safety assessment of a rock-cavern type LILW repository

This study presents a practical application of complementary safety indicators, which can be applied in a safety assessment of a radioactive waste repository by excluding a biosphere simulation and comparing the artificial radiation originating from the repository with the background natural radiation. Complementary safety indicators (radiotoxicity flux from geosphere and radiotoxicity concentration in seawater) were applied in the safety assessment of a rock-cavern type low and intermediate level radioactive waste (LILW) repository in the Republic of Korea. The natural radionuclide (40K, 226,228Ra, 232Th, and 234,235,238U) concentrations in the groundwater and seawater at the Gyeongju LILW repository site were measured. Based on the analyzed concentrations of natural radionuclides, the levels of natural radiation were determined to be 8.6 × 10−5 – 8.0 × 10−4 Sv/m2/yr and 6.95 × 10−5 Sv/m3 for radiotoxicity flux from the geosphere and radiotoxicity concentration in seawater, respectively. From simulation results obtained using a Goldsim-based safety assessment model, it was determined that the radiotoxicity of radionuclides released from the repository is lower than that of the natural radionuclides inherently present in the natural waters. The applicability of the complementary safety indicators to the safety case was discussed with regard to reduction of the uncertainty associated with biosphere simulations, and communication with the public.

PRELIMINARY DESIGN OF POTENTIAL STORAGE FACILITY FOR LOW AND INTERMEDIATE LEVEL RADIOACTIVE WASTE

An issue concerning the management of radioactive waste (RAW) in the Republic of Croatia continuously rises due to the lack of systematic solutions. In this paper, emphasis is being placed on the storage of RAW, as one of the important phases of RAW management. The current RAW storage facilities are not sufficient and there is a need for a new facility for low and intermediate level radioactive waste (LILW), since such waste accounts for the largest part of RAW. For this reason, a preliminary estimate for a one-storey structure was carried out, a potential storage facility of LILW that would be constructed at the location of 􀀬erkezovac, Trgovska Gora, in the Republic of Croatia, as the currently most prominent location for the realization of the RAW storage facility. The geological characteristics of the site were analysed and, in accordance with the relevant regulations, a static design of the facility was made in the ETABS and SAFE program packages. The construction is a typical reinforced concrete structure, rectangular in plan (10.0 x 30.0 m) with a height of 5.9 m. The selected location meets the given parameters of geological characteristics, while the construction meets the given load and serviceability conditions de􀆤 ned by Eurocode (ultimate limit state and serviceability limit state). However, a more detailed analysis would be necessary if the decision to construct such a storage facility would be made.

An Experimental Study on Treatment of Typical Low and Intermediate Level Radioactive Wastes with Thermal Plasma Melting Technology

With the active development of the nuclear industry in China, it is extremely urgent to study and develop advanced technologies for the radioactive waste treatment and disposal. With features of high temperature, high energy density and broad applicability, the thermal plasma melting technology has been considered as one of the key subjects in the radioactive waste treatment research field. A pilot scale experimental study was carried out on the treatment of typical low and intermediate level radioactive wastes (LILWs) from the nuclear power plant in China with thermal plasma melting technology. Two representative wastes, thermal insulation waste and cotton, were selected and melted in the plasma melting furnance at 1250°C for ~ 100 minutes until they were completely vitrified. Meanwhile, tracers of Co2O3, CsCl and SrCO3 were added to the original wastes to simulate the containment effect of 58Co and 60Co, 134Cs and 137Cs as well as 90Sr in the vitrification during the treatment process. The XRD patterns of both vitrification samples displayed as the typical amorphous state. Performances of resultant glass waste-forms, including the physical property, the leachability and the mechanical capacity, were in accordance with the standard requirements of the hign level radioactive waste glass vitrificaton, and were better than those of the cement solidified waste. Further experimental work is in need to better understand the radioactive nuclide migration regularity and to improve the performance of the plasma melting system.

Uncertainty Management on Human Intrusion Scenario Assessment of the Near Surface Disposal Facility for Low and Intermediate-Level Radioactive Waste: Comparative Analysis of RESRAD and GENII

In order to manage the uncertainty about the evaluation and analysis of the human intrusion scenario of the Gyeongju Low and Intermediate Level Radioactive Waste(LILW) disposal facility, the calculation result by the GENII code was assessed using the RESRAD code, which was developed to evaluate the radiation effects of contaminated soil. The post-drilling scenario was selected as a human intrusion scenario into the near-surface disposal facility to analyze the uncertainty of the modeling by identifying any limitations in the simulation of each code and comparing the evaluation results under the same input data conditions. The results revealed a difference in the migration of some nuclides between the codes, but confirmed that the dose trends at the end of the post-closure control period were similar for all exposure pathways. Based on the results of the dose evaluation predicted by RESRAD, sensitivity analysis on the input factors was performed and major input factors were derived. The uncertainty of the modeling results and the input factors were analyzed and the reliability of the safety evaluation results was confirmed. The results of this study can be applied to the implementation ‘Safety Case Program’ for the Gyeongju LILW disposal facility.

The Status and Experiences of LILW Disposal Facility Construction

This paper presents the status and experiences of Low and Intermediate-level radioactive Waste(LILW) disposal facility in Korea. The LILW disposal facility of 800,000 drums (200 L/drum) is being constructed in the site of KORAD's LILW Disposal Center(206 m2, Gyeongju-si). The first stage of the LILW facility construction was completed in June 2014, and approved for use by Nuclear Safety and Security Commission (NSSC) in December 2014. The first stage of the LILW facility is a type of underground silo with disposal capacity of 35,200 m3 (100,000 drums). The second stage of the LILW facility is under construction to be completed by December 2020. Site construction is currently underway. The second stage of the LILW facility is a type of near-surface disposal with disposal capacity of 32,875 m3 (125,000 drums)

Assessment of strontium transport in clay medium by using a reactive transport modelling approach – an HP1 case study model

Modelling of radionuclide transport in the geosphere is a key issue of the evaluation of safety of low- and intermediate-level radioactive waste (LILW) repositories. In Bulgaria, due to the operation of Kozloduy Nuclear Power Plant (NPP), a building of such facility is ongoing. The migration of 90Sr was evaluated by simulation of flow and chemical modelling, performed with coupled numerical code HP1. The adsorption properties of the clays were implemented in the code by surface complexation reactions between strontium and iron oxides. The results were compared with another model performed with computer code HYDRUS 1-D, in which the adsorption properties of the medium were evaluated with the relevant for strontium distribution coefficient, Kd.

폐밀봉선원 처분시스템 예비 폐쇄후 안전성평가

An optimum disposal plan of disused sealed radioactive sources (DSRSs) should be established to ensure long-term disposal safety at the low- and intermediate-level radioactive waste (LILW) disposal facility in Gyeongju. In this study, an optimum disposal system was suggested and preliminary post-closure safety assessment was performed. The DSRSs disposal system was composed of a rock cavern and near surface disposal facilities at the Gyeongju LILW disposal facility. The assessment was conducted using GoldSim program, and probabilistic assessment and sensitivity analysis were implemented to evaluate the uncertainties in the input parameters of natural barriers. Deterministic and probabilistic calculations indicated that the maximum dose was below the regulatory limits (<TEX>$0.1mSvyr^{-1}$</TEX> for the normal scenario, <TEX>$1mSvyr^{-1}$</TEX> for the well scenario). It was concluded that the DSRSs disposal system would maintain environmental safety over a long-time. Moreover, the partition coefficient of Np in host rock, Darcy velocity in host rock, and density of the host rock were the most sensitive parameters in predicting exposure dose in the safety assessment.

Post-closure safety assessment of near surface disposal facilities for disused sealed radioactive sources

Great attention has been recently paid to the post-closure safety assessment of low- and intermediate-level radioactive waste (LILW) disposal facility for disused sealed radioactive sources (DSRSs) around the world. Although the amount of volume of DSRSs generated from industry, medicine and research and education organization was relatively small compared with radioactive wastes from commercial nuclear power plants, some DSRSs can pose a significant hazard to human health due to their high activities and long half-lives, if not appropriately managed and disposed. In this study, post-closure safety assessment was carried out for DSRSs generated from 1991 to 2014 in Korea in order to ensure long-term safety of near surface disposal facilities. Two kinds of disposal options were considered, i.e., engineered vault type disposal facility and rock-cavern type disposal facility. Rock-cavern type disposal facility has been under operation in Gyeongju city, republic of Korea since August 2015 and engineered vault type disposal facility will be constructed until December 2020 in the vicinity of rock-cavern disposal facility. Assessment endpoint was individual dose to the member of critical group, which was modeled by GoldSim, which has been widely used as probabilistic risk analysis software based on Monte Carlo simulation in the area of safety assessment of radioactive waste facilities. In normal groundwater scenario, the maximum exposure dose was extremely low, approximately 1×10−7mSv/yr, for both disposal options and satisfied the regulatory limit of 0.1mSv/yr. However, in the drinking well scenario, the maximum exposure dose for engineered vault type disposal facility was assessed as 2.022mSv/yr where the value exceeded the regulatory limit of 1mSv/yr. The maximum exposure dose for rock-cavern type disposal facility was calculated to be 0.634mSv/yr, whose results was relatively very close to the regulatory limit considering high uncertainty of long-term environmental conditions. It was demonstrated that DSRSs including the radionuclides of 14C, 226Ra and 241Am significantly contributed to the large portion of exposure dose to the public based on the long-term safety assessment. Therefore, it was recommended that the near surface disposal of DSRSs containing 14C, 226Ra and 241Am should be restricted and managed by long-term interim storage option in order to minimize their potential radiological health effects.