Nuclide release from an HLW repository: Development of a compartment model

A compartment model, ACGEO, has been developed for the Korean reference Repository System (KRS) for an HLW disposal by utilizing AMBER, a general compartment modeling tool. ACGEO is a flexible and adaptable tool, by which a transient calculation of both a nuclide release and transport in the near- and the far-field of a repository system with various and complex shapes can be carried out. It is expected to be used for the treatment of nuclide transport of a decay chain for safety assessment, as well as for design feedback on the KRS both on a deterministic and a probabilistic bases. Some illustrative cases are also investigated in which the influence of varying the near-field system features on the release of selected nuclides for an HLW repository are examined, in order to demonstrate the usability of ACGEO for feedback to a refined repository concept and for establishing an appropriate safety margin for an HLW repository.     

Monte Carlo estimation of radionuclide release at a repository scale

Prediction of radionuclide release is a central issue in the performance assessment of nuclear waste repositories. This requires modeling of the radionuclide migration processes through the repository barriers, accounting for the related uncertainties. The present paper illustrates a Monte Carlo simulation-based compartment model in which detailed, local-scale modeling feeds a global-scale analysis of the repository, at reasonable computational expenses. An application to a realistic case study is presented to verify the feasibility of the approach.     

Migration processes in a model deep repository for the disposal of intermediate-level nuclear waste

It is difficult to conceive of radionuclides escaping from a repository by any means other than migration in groundwater. Simple models of the repository are constructed and various migration processes are identified and assessed, according to the flow speed of water through the repository. Diffusion in static water and advection in fast flows are considered separately initially, but later we examine the effect of slow flows in which both these processes contribute to the removal of radionuclides. Concentration profiles across the repository, fluxes of nuclides and total losses are obtained from the analysis. We investigate the time scales necessary for the steady state to be achieved in the repository and conclude that flow speed is roughly inversely proportional to this time scale, i.e. faster flows establish a steady state sooner than slow ones. We also assess the sensitivity of the results to the physical properties of the components of the repository.     

The transport of radionuclides discharged from a geologic waste repository

A 3-D conservative solution to the convection-dispersion-retardation equation describing the migration of radionuclides from a geologic nuclear waste disposal site is derived using a Green's function approach under conditions in which the contaminant inventory is preferentially discharged to the steady groundwater flow. The numerical implementation of the upper estimate to the transported inventory, as embodied in the CANUKE© code, is described; its applicability is demonstrated by calculating the groundwater concentration of the neptunium series nuclides released from used CANDU™ fuel emplaced in an hypothetical repository; and its general usage is discussed. The ability to handle actinide chains of arbitrary length, whatever their initial inventories, yields both less conservative and physically more realistic estimates of transported inventories than schemes which begin with a contracted chain of longer-lived nuclides whose progeny are assumed to accompany them in their migration, irrespective of their relative mobilities.     

放射性核素量热计研制

量热法可实现放射性核素质量的非破坏性直接测量,是核材料计量最为便捷和准确的方法之一.为实现特定包容体尺寸及热功率范围核材料样品的高效测量,设计研制了热流型双杯量热计.通过甄选精密仪表部件,关键温差传感部件的结构优化以及合理选型制作,结合自动化测控方法,解决了量热测量关键技术问题,研制出适用于瓦功率量级的放射性核素量热计...     

Simple methods of including removal processes in models of the long-range transport of airborne material

Some simple methods are suggested for correcting models of the dispersion of airborne material over long distances so as to include the effect of dry and wet deposition. The methods are no more complicated than conventional methods for treating the continuous release of radioactive materials, but have advantages in treating conditions more realistically.     

Probability methods in homogenization schemes for transport processes

Homogenization schemes (small mean free path or diffusion limit) for probabilistic transport processes are considered with attention focused on the underlying boundary layer processes. Some of the hypotheses and constructions necessary for the homogenization expansion are analyzed using the theory of ergodic Markov processes. In particular, Fredholm alternative type results and exponential decay estimates are obtained for the leading order operators in the expansion, thus justifying certain of the matching operations between the interior expansions and the boundary layer expansions.     

Cementitious backfill materials for a L/ILW repository - Investigations of gas transport properties

Significant gas production is expected in low- and intermediate-level waste (L/ILW) repositories. The backfill material surrounding the waste containers must therefore — besides other requirements — ensure that this gas can leave the repository near-field. A wide selection of cementitious materials were studied with respect to the requirements. Because the permeability depends on the total pore volume as well as on the pore size distribution, both these parameters were varied by changing the composition of the matrix, and the type and grain size distribution of the aggregates. The permeability was determined on specimens which had been cured under various storing conditions for one and three months. As expected the results show that permeability depends on the conditions and the duration of the curing. For water-saturated specimens a threshold pressure can be observed below which no gas flow occurs.     

Monte Carlo simulation of radionuclide transport through fractured media

The presence of fractures within the host matrix may strongly affect the radionuclide mobility in the far-field of the current proposed solutions for underground radioactive waste repositories. In this paper a Monte Carlo simulation approach, based on analogies with neutron transport, has been developed for describing radionuclide transport in fractured media. The Monte Carlo approach here adopted is verified by means of comparison with analytic and numerical solutions for two different case studies. An application to a more complex and realistic case is also reported to show the potential of the proposed approach.     

手持式快速核素检测仪的研制

多用途手持式核素检测仪以基于ARM7的多道分析器为核心,可选配电致冷CdTe半导体探测器或闪烁计数器.内嵌常见的天然和人工放射线核素库,可实现现场原位测量和核素检测.实验表明,CdTe探测器对59.54和0.662 keVγ射线的能量分辨率分别可达0.8和5.9 keV,能完成中、低能核素的检测.闪烁计数器可实现中、高...     

A probabilistic study of the effect of retardation factor uncertainty using a compartment model for radionuclide release into the biosphere

The effect of retardation factor uncertainty with reference to the results of a compartment model, representing radionuclide release into the biosphere from a disposal facility, is studied through the application of polynomial chaos theory. We review the derivation of the working equations and apply the polynomial chaos expansion to these equations. The uncertainty in the retardation factor typically covers several orders of magnitude and stems from uncertainties in the distribution coefficient, these large uncertainties make any quantitative analysis of radionuclide release highly problematic. In this paper, we assume that the uncertainty in the retardation factor is fully described by a log-uniform distribution. We compare results using polynomial chaos against a semi-analytical solution for a short decay chain and present numerical results from polynomial chaos applied to a longer generic decay chain.     

Integral transport in a three region sphere with associated benchmark problems

We derive, by the method of characteristics, an integral form of the transport equation with isotropic scattering for spherical symmetry in which the spatial dependence of the cross-sections is arbitrary. The sources of particles (photons or neutrons) can be either incident on the surface with an arbitrary angular distribution or distributed within the body. A specific example is given for the case of a three region sphere in the form of a central spherical core surrounded by a concentric shell followed by an outer shell. By allowing the total cross-section to go to zero in any one of the regions we may evaluate the effect of voids. Some numerical results are given which enable the limitations of void representation in a standard computer code to be highlighted. We also consider the case of a multiplying system and calculate the effect of an annular void on the multiplication factor.     

Analytic solutions of radionuclide transport with the limited diffusion from the fracture into a porous rock matrix

To identify the effect of the limiting zone for the matrix diffusion from an open fracture into surrounding medium recently found from field studies, full analytic solutions are derived. The solutions are then verified with existing solutions and a numerical inversion of the Laplace transform approach. The relevant field data are collected and then numerical evaluations are pursued for these complicated analytic solutions with multiple integrals. Numerical results illustrate that for given data sets the effect of limiting zone for the rock matrix diffusion can be negligible so that the current infinite rock model may underestimate the radionuclide transport through fractured porous rock.     

Multigroup multi-layer models of neutron reflection and transmission for reactor transport calculations with anisotropic scattering

In this article, we extend the one-speed multi-layer models to neutron reflection and transmission developed in our earlier work (de Abreu, M.P., 2005. Multi-layer models to neutron reflection and transmission for whole-core transport calculations, Annals of Nuclear Energy 32, 215) to multigroup transport theory. We begin by considering a two-layer boundary region, and we develop for such a region discrete ordinates models to the diffuse reflection and transmission of neutrons for multigroup nuclear reactor core problems with anisotropic scattering. We perform numerical experiments to show that our models to neutron reflection and transmission can be used to replace efficiently and accurately two nonactive boundary layers in whole-core transport calculations. We conclude this article with an inductive extension of our two-layer results to a boundary region with an arbitrary number of layers.     

Estimation of the inventory of the radioactive wastes in Fukushima Daiichi NPS with a radionuclide transport model in the contaminated water

Quantification of the radioactive waste inventory remaining inside the reactors at Fukushima Daiichi NPS is necessary to effectively plan their recovery, treatment, and disposal. Analysis of radionuclide concentrations and secondary wastes in the contaminated water treatment system can provide a means to estimate the radioactive waste inventory, which is not possible by more direct methods due to problems of accessibility and high levels of radiation. A predictive model has therefore been developed to estimate the radioactive waste inventory from the radionuclide concentrations and throughputs in the contaminated water. Model fitting has enabled the estimation of the key parameters, such as the initial radionuclide concentration C₀, the continuous release rate F, and inventory of source of continuous release I_S0. An estimated one-third of the total ¹³⁷Cs inventory has already made its way into the water treatment system as secondary wastes, whereas half still remains inside the damaged reactors as of 13 March 2014.     

Progress of a cost optimization for an HLW repository in Korea

This paper provides the results of a cost optimization for a CANDU spent fuel canister as well as the operational duration of an HLW repository. From the design change of an advanced-CANDU spent fuel canister, the overall costs were expected to be reduced by 124 MEUR in the case of disposing of 36,000 tU in an HLW repository, and it was also found that the optimal operational duration for an HLW repository was 83 years, to minimize the total cost. But this operational duration was only calculated from the aspect of cost benefits with economics' perspectives.We confirmed that the canister and operational duration are the dominant cost drivers for surface facilities and underground facilities for a cost optimization, respectively. Especially, the manufacturing method of an outer canister using the cold spray coating technique which was developed through collaboration with a domestic company is suggested to minimize the overall costs.