Very Low Level Waste Research Articles

A study on the effect of material impurity concentration on radioactive waste levels for plans for decommissioning of nuclear power plant

Co and Eu impurities in the SSCs are nuclides that dominantly influence the neutron-induced radioactive inventory in metal and concrete radwastes (radioactive wastes) during NPP decommission. The impurity concentrations provided by NUREG/CR-3474 were used for the practical range of Co and Eu impurity concentrations to be applied to the code calculations. Metal structures near the core were evaluated to be ILW (intermediate-level waste) for the whole range of Co impurity concentration, so the boundary line between ILW and LLW (low-level waste) has no change for the whole concentration range provided by NUREG/CR-3474. Also, the boundary line between VLLW (very low-level waste) and CW (clearance waste) in the concrete shield could alter a little depending on the Eu impurity concentration within the range provided by NUREG/CR-3474. From this work, it is found that the concentration of material impurities of SSCs gives no critical impact on determining radwaste levels.

Mechanisms Governing 90Sr Removal and Remobilisation in a VLLW Surface Disposal Concept

Flow-through columns were used to assess potential long-term trends in 90Sr biogeochemistry and transport in a Finnish near-surface very low-level waste (VLLW) repository concept. Experiments simulated the effects of water intrusion and flow through the repository barrier and backfill materials, examining impacts on 90Sr migration. Artificial rainwater containing 2.0 mg/L stable Sr (as a proxy for 90Sr) was pumped through column systems that had varying compositions from a matrix of rock flour (backfill material), bentonite (backfill/sealing material), and carbon steel (waste encapsulation material), for 295 days. Effluent geochemistry was monitored throughout. Sr retention behaviour in all column systems was broadly similar. Sr removal from influent rainwater was marked (~95% removed) at the beginning of the experiments, and this degree of removal was maintained for 20 days. Thereafter, Sr concentrations in the effluents began to rise, reaching ~2 mg/L by 295 days. Further, 56%–67% of added Sr was retained in the repository materials over the 295-day reaction period. Analysis of the effluents indicated that colloids did not form; as such, Sr output was likely to be aqueous Sr2+. Upon completion of the experiment, solid-associated Sr distribution and speciation in the columns were assessed through column sectioning and post-mortem analyses, which encompassed the following: total acid digests, sequential extractions, and XAS analysis. The total acid digests and sequential extractions showed that Sr was evenly distributed throughout the columns and that the majority (68%–87%) of solid-associated Sr was in the exchangeable fraction (MgCl2). This suggested that a major part of the solid-phase Sr was weakly bound to the column materials via outer-sphere sorption. Interestingly, a smaller amount of Sr (7%–23%) could only be extracted by aqua regia, suggesting that a proportion of Sr may bind more strongly to the barrier materials. XAS analysis of select samples confirmed that the dominant Sr phase was sorbed to the rock flour and bentonite, but not corroded carbon steel. Columns were also subject to remobilisation experiments using artificial rain- and seawater without added Sr. While rainwater remobilised Sr slowly, high-ionic strength seawater remobilised Sr at much higher rates in the systems containing bentonite. Interestingly, Sr was well retained in the rock flour-only system following rain and seawater intrusion. Overall, the results indicate that the column materials provide reactive surfaces for Sr removal should it be released from waste packages; however, the backfill and barrier materials have limited retention capacity, and the dominant sorption interaction is relatively weak. The safety case for the shallow disposal of radioactive waste should consider the possibility of seawater intrusion and that the bentonite-bound Sr was significantly more susceptible to remobilisation following seawater, despite retaining slightly more Sr during sorption experiments.

Study on Dose Rate on the Surface of Cask Packed with Activated Cut-off Pieces from Decommissioned Nuclear Power Plant

Background: Reactor pressure vessel (RV) with internals (RVI) are activated structures by neutron irradiation and volume contaminated wastes. Thus, to develop safe and optimized disposal plan for them at a disposal site, it is important to perform exact activation calculation and evaluate the dose rate on the surface of casks which contain cut-off pieces. Materials and Methods: RV and RVI are subjected to neutron activation calculation via Monte Carlo methodology with MCNP6 and ORIGEN-S program—neutron flux, isotopic specific activity, and gamma spectrum calculation on each component of RV and RVI, and dose rate evaluation with MCNP6. Results and Discussion: Through neutron activation analysis, dose rate is evaluated for the casks containing cut-off pieces produced from decommissioned RV and RVI. For RV cut-off ones, the highest value of dose rate on the surface of cask is 6.97 × 10-1 mSv/hr and 2 m from it is 3.03 × 10-2 mSv/hr. For RVI cut-off ones, on the surface of it is 0.166 × 10-1 mSv/hr and 2 m from it is 1.04 × 10-1 mSv/hr. Dose rates for various RV and RVI cut-off pieces distributed lower than the limit except the one of 2 m from the cask surface of RVI. It needs to adjust contents in cask which carries highly radioactive components in order to decrease thickness of cask. Conclusion: Two types of casks are considered in this paper: box type for very-low-level waste (VLLW) as well as low-level waste (LLW) and cylinder type for intermediate-level waste (ILW). The results will contribute to the development of optimal loading plans for RV and RVI cut-off pieces during the decommissioning of nuclear power plant that can be used to prepare radioactive waste disposal plans for the different types of wastes—ILW, LLW, and VLLW.

Scenarios and Pathways of Radionuclide Releases from Near-Surface Waste Disposal Facilities: A Brief Overview of Historical Evidence

The very low-level waste (VLLW) produced during decommissioning of nuclear facilities can be suitable for disposal in landfill type facilities. Considering the similarities in design, the experience gained in near-surface disposal of radioactive waste in trenches and vaults is relevant to the issue of VLLW disposal in landfills. This paper presents a brief review of internationally reported cases of radionuclide releases from near-surface disposal facilities. Based on this review, the conclusions are made that the following radionuclide release and exposure scenarios should be accounted for in safety assessment of VLLW disposal in landfills: i) leaching from waste to groundwater by atmospheric precipitations; ii) bath-tubing scenario; iii) scenarios caused by extreme meteorological and hydrological events (erosion, flooding, etc.); iv) human intrusion. The gaseous transport deserves attention for a number of relevant radionuclides, such as (C-14, Rn-222, etc.). In addition, the possibility of early degradation of engineered containment structures (soil covers, bottom seals) should be cautiously considered.

A Study on Segmentation Process of the K1 Reactor Vessel and Internals

After the permanent shutdown of K1 in 2017, decommissioning processes have attracted great attention. According to the current decommissioning roadmap, the dismantling of the activated components of K1 may start in 2026, following the removal of its spent fuel. Since the reactor vessel (RV) and reactor vessel internal (RVI) of K1 contain massive components and are relatively highly activated, their decommissioning process should be conducted carefully in terms of radiological and industrial safety. For achieving maximum efficiency of nuclear waste management processes for K1, we present activation analysis of the segmentation process and waste classification of the RV and RVI components of K1. For RVI, the active fuel regions and some parts of the upper and lower active regions are classified as intermediate-level waste (ILW), while other components are classified as low-level waste (LLW). Due to the RVI’s complex structure and high activation, we suggest various underwater segmentation techniques which are expected to reduce radiation exposure and generate approximately nine ILW and nineteen very low level waste (VLLW)/LLW packages. For RV, the active fuel region and other components are classified as LLW, VLLW, and clearance waste (CW). In this case, we suggest in-situ remote segmentation in air, which is expected to generate approximately forty-two VLLW/LLW packages.

Disposal of very low-level radioactive waste: Lithuanian case on the approach and long-term safety aspects

Decommissioning of nuclear power plants typically requires management of hundreds of thousands of cubic meters of waste with very low levels of radioactivity. Such very low-level waste (VLLW) is suitable for disposal in simple landfill type facilities.The only nuclear power plant in the Republic of Lithuania, the Ignalina NPP, was finally shut down at the end of 2009. Final management of the existing operational radioactive waste and that arising during the decommissioning will require the disposal of 60,000 m3 of very low-level short-lived solid waste. An overview of the implementation of the landfill disposal facility for very low-level radioactive waste at the Ignalina NPP site and the main aspects of long-term safety assessments based on national regulations as well as the IAEA's recommendations including the siting process are presented in this paper as well as a short description of the disposal concept and the characteristics of the waste intended for disposal in the landfill. Water pathway scenarios as well as intrusion scenarios have been investigated following an IAEA methodology. It has been defined that 14C is the greatest contributor to the total dose in the case of water pathway scenarios. In the case of intrusion scenarios, the main contribution to the total dose is from radionuclides 94Nb and 137Cs. In all the considered cases, the total value of the annual effective exposure dose remains quite well (orders of magnitude) below the value of the dose constraint established by regulations of Lithuania. Uncertainty analysis is carried out to define the influence of various factors on the obtained results of the safety assessment.

Modelling contaminant transport in generic landfills for decommissioning waste from German nuclear power plants

The German decision to phase out nuclear power will increase the quantity of very low level waste (VLLW) from nuclear power plant (NPP) decommissioning in the near future, which will cause a capacity shortage of landfills. As a likely consequence, landfills will have to be constructed that are exclusively built for decommissioning waste without addition of municipal waste, and assessments for such types of landfills are not available in Germany.The present study investigates flow and transport processes in generic landfills that only contain decommissioning waste. The generic landfills comply with German technical standards such as the Landfill Ordinance. Simulations of contaminant transport are performed using the codes SiWaPro DSS and SPRING. The simulations show that Sr-90 is the first radionuclide that leaves the landfill due to its low sorption in the waste body. In contrast, no release of Cs-134, Cs-137 and Am-241, which have higher sorption coefficients, was simulated. Our work also shows that three-dimensional modelling of flow and transport is required if the landfill has a lateral drainage. Results suggest that the migration of radionuclides into the environment could be reduced by installing a drainage layer with high sorption capacity on top of the landfill's bottom seal. Sorption of Sr-90 within this drainage layer would delay the release of Sr-90 from the landfill, thereby giving more time for the Sr-90 decay inside the landfill.

Geochemical Study of Engineering Shield of Uranium, Strontium and Cesium at a very Low Level Waste Disposal Site

Advantages and disadvantages of natural shields in a very low level waste (VLLW) disposal area were studied with regard to the adsorption and fixation capability of the natural shields against radioactive elements Uranium, Strontium and Cesium, with the understanding of the hydrogeological and geochemical characteristics of the surrounding area. Necessity of establishing geochemical engineering shield in the disposal area is discussed based on the understanding of the inadequacy of the natural shields against Uranium and Strontium. A physical chemistry model involving three-fold of shields suitable for the disposal area is proposed with integrated consideration of the results from efficacy and endurance experiments on the natural shield.

Multi-parameter Experimental Adsorption Effect of 239Pu on Phyllitic Slate and Clay

The paper studies the adsorption features of 239Pu in phyllitic slate and clay media through static experiment method. The two different geo-media were collected from a site for disposing very low level waste (VLLW) as experimental samples. Experimental results have shown that an adsorption equilibrium time of 239Pu in phyllitic slate and clay is approximately 10 days, and the adsorption coefficient (Kd) B Bvalues of phyllitic slate and clay are about 6000ml/g and 2700ml/g, respectively. The experimental result of particle size effect indicates when adsorption of plutonium in particle samples increases, the KBdB values of these two geo-media increase while the particles decrease in size. And the experimental impact on aqueous phase pH value reveals that Kd of 239Pu in creases at higher aqueous phase pH value, illustrating stronger adsorption capacity. Experimental data are also measured through the curve fitting, KBdB values are thus obtained theoretically.

중국에서의 극저준위 방사성 고체 폐기물 관리

This paper introduces the policy and regulations on very low level waste (VLLW) management in China. Given the important decommissioning and site restoration program of the old facility, it is considered necessary to create a new disposal facility dedicated to VLLW. Many general design principles are in common with to the disposal facility for low and intermediate level waste (LILW), namely the isolation of the waste by means of a multibarrier system, but using bentonite and/or high density polyethylene membranes instead of the generalized use of concrete barriers. The design of the facility is consistent with the design of disposal facilities for hazardous waste. The engineering design of two VLLW disposal facilities is introduced.

Sorption of strontium and fractal scaling of the heterogeneous media in a candidate VLLW disposal site

Because of the deposit and accumulation from the debris flow, the heterogeneous geological characteris- tics is obvious for a candidate very low level waste (VLLW) disposal site, with the grain size ranging from tens of microns to 75 cm. Therefore, it is challenging to directly measure the sorption capacity of the media and the distri- bution coefficient of some radionuclides, such as strontium. We have studies the correlation of the particle mass content with different grade size and the sorption capacity, which is important in the modeling of radionuclide migration in the heterogeneous disposal site. A total of three deep pits and five shallow trenches were excavated, and 21 solid samples were collected for laboratory experiments. The grade and percentage of the different-sized particles were obtained, and the fractal dimension (D) of the media was calculated from the results of sieved experiments. Steady state sorp- tion time and sorption isotherm of strontium was deter- mined in the heterogeneous media, and sorption and distribution of strontium in the heterogeneous media were evaluated by the relationship between the mass percentage and distribution coefficient (Kd) of the fine-particle media, which was comprised of selected particles with a diameter less than 1 mm, and the correlation on the Kd and D was regressed fit. The results indicated that fractal dimension bounded from 2.39 to 2.62 in the media, and Kd values of strontium ranged between 119 and 126 in the fine-particle media, and corresponding value was 11 and 43 in the ori- ginal media. The correlation between Kd and D was approximately linear.

Sorption and retardation of strontium in fine-particle media from a VLLW disposal site

Study of fine-particle media, because of their high sorption capacities, is of particular importance for the use as backfill materials in waste repository design, and because argillaceous formations are particularly suitable as host rock formations. In this study, sorption and retardation characteristics of strontium in fine-particle media were studied to evaluate the distribution coefficient (Kd) and retardation factor (Rd) of this radioactive element in fine-particle media, which was comprised of selected particles with a diameter less than 1 mm from a candidate site to dispose very low level waste (VLLW). The results indicated that Kd values of strontium under different initial concentrations ranged between 20 and 110. Values of strontium Rd measured from column experiments ranged between 36 and 102, with the corresponding Kd values, determined from solving the inverse problem of Rd calculating formula, ranging between 5 and 20. In conclusion, the Kd value of Sr from the batch tests was found to be higher than these from the column experiments.

Preliminary safety concept for disposal of the very low level radioactive waste in Romania

In Romania, there are certain nuclear installations in operation or under decommissioning, all of them representing an important source of very low level waste (VLLW). This paper presents an overview on the approach of the VLLW management in Romania, focused on those resulted from the nuclear power plants decommissioning. At the same time, the basic elements of safety concept, together with some safety evaluations concerning VLLW repository are presented and discussed too.

Radioactive waste disposal implications of extending Part IIA of the EnvironmentalProtection Act to cover radioactively contaminated land

A short study has been carried out of the potential radioactive waste disposal issuesassociated with the proposed extension of Part IIA of the Environmental Protection Act1990 to include radioactively contaminated land, where there is no other suitable existinglegislation.It was found that there is likely to be an availability problem with respect todisposal at landfills of the radioactive wastes arising from remediation. This isexpected to be principally wastes of high volume and low activity (categorisedas low level waste (LLW) and very low level waste (VLLW)). The availabilityproblem results from a lack of applications by landfill operators for authorisation toaccept LLW wastes for disposal. This is apparently due to perceived adversepublicity associated with the consultation process for authorisation coupled withuncertainty over future liabilities. Disposal of waste as VLLW is limited both byquestions over volumes that may be acceptable and, more fundamentally, by thelikely alpha activity of wastes (originating from radium and thorium operations).Authorised on-site disposal has had little attention in policy and guidance in recent years,but may have a part to play, especially if considered commercially attractive.Disposal at BNFL’s near surface disposal facility for LLW at Drigg is limited to wastesfor which there are no practical alternative disposal options (and preference hasbeen given to operational type wastes). Therefore, wastes from the radioactivelycontaminated land (RCL) regime are not obviously attractive for disposal to Drigg.Illustrative calculations have been performed based on possible volumes and activitiesof RCL arisings (and assuming Drigg’s future volumetric disposal capacity is950 000 m3). These suggest that wastes arising from implementing the RCL regime, if all disposed toDrigg, would not represent a significant fraction of the volumetric capacity of Drigg,but could have a significant impact on the radiological capacity with respect to226Ra plus232Th.The government’s decision-making programme for managing solid radioactive wastes in theUK may possibly achieve a general consensus that the use of landfill for LLW from theRCL regime has a fundamental role to play. However, this is unlikely to change thesituation within the next few years. No new national facility arising from this programme islikely to be available during the first decade of the operation of a new RCL regime.Hence it appears that Drigg will need to play an important role for some years tocome.

Évaluation de l'impact radiologique de différentes voies de gestion des déchets TFA

Radiological impact assessment of very low-level radioactive waste management pathways. Two approaches are currently being adopted for the management of very low-level waste (VLLW): disposal in surface facilities and, for some categories of VLL waste, the implementation of treatment and utilisation channels. In this paper, the orders of magnitude of the radiological consequences which may result from these two options are discussed, in a matter of a description rather than a comparison. As regards disposal, radiological impact is presented through a generic case, along with the uncertainties stemming from the variability of the parameters associated with its design, as well as those associated with the geosphere and the biosphere. In the case of treatment and utilisation channels, such as disposal or recycling of VLL waste oils or VLL wastes containing asbestos, the assessment of radiological consequences deals with the public exposure and with the occupational exposure on the basis of a stage-by-stage breakdown of the industrial process.