ISO Standardization of the Scaling Factor Method for Low- and Intermediate Level Radioactive Wastes Generated at Nuclear Power Plants

Low- and intermediate-level radioactive wastes (L-ILW) generated at nuclear power plants are disposed of in various countries. In the disposal of such wastes, it is required that the radioactivity concentrations of waste packages should be declared with respect to difficult-to-measure nuclides (DTM nuclides), such as C-14, Ni-63 and α-emitting nuclides, which are often limited to maximum values in disposal licenses, safety cases and/or regulations for maximum radioactive concentrations. To fulfill this requirement, the Scaling Factor method (SF method) has been applied in various countries as a principal method for determining the concentrations of DTM nuclides. In the SF method, the concentrations of DTM nuclides are determined by multiplying the concentrations of certain key nuclides by SF values (the determined ratios of radioactive concentration between DTM nuclides and those key nuclides). The SF values used as conversion factors are determined from the correlation between DTM nuclides and key nuclides such as Co-60. The concentrations of key nuclides are determined by γ ray measurements which can be made comparatively easily from outside the waste package. The SF values are calculated based on the data obtained from the radiochemical analysis of waste samples. The use of SFs, which are empirically based on analytical data, has become established as a widely recognized “de facto standard”. A number of countries have independently collected nuclide data by analysis over many years and each has developed its own SF method, but all the SF methods that have been adopted are similar. The project team for standardization had been organized for establishing this SF method as a “de jure standard” in the international standardization system of the International Organization for Standardization (ISO). The project team for standardization has advanced the standardization through technical studies, based upon each country’s study results and analysis data. The conclusions reached by the project team was published as ISO International Standard 21238:2007 “The Scaling Factor method to determine the radioactivity of low- and intermediate-level radioactive waste packages generated at nuclear power plants” [1]. This paper gives an introduction to the international standardization process for the SF method and the contents of the recently published International Standard.

Characteristics of Solidified Products Containing Radioactive Molten Salt Waste

For a treatment of molten salt wastes generated from a pyroprocessing of oxide spent fuel, we had suggested a stable chemical route, named GRSS (Gel-Route Stabilization & Solidification), and a subsequent consolidation method. By using this method, a series of monolithic wasteforms with different conditions were fabricated, and then their physicochemical properties were investigated. A simulated salt containing 90wt% LiCl, 6.8wt% CsCl, and 3.2wt% SrCl2 was treated with a gel-forming material system, Si/Al/P = 0.4/0.4/0.2 and 0.35/0.35/0.3, and the gel-products were treated at 1100C° after mixing with borosilicate glass powder, where the salt loadings were about 16∼20wt%. The solidified products had a density of 2.3∼2.35g/cm3, a micro-hardness of 4.69∼4.72GPa, a glass transition temperature of 528∼537C°, and a thermal expansion coefficient of 1.65×10−7∼3.38×10−5/C°. Leaching results by the PCT-A method revealed leached rates, 10−3∼10−2g/m2day and 10−4∼10−3g/m2day for Cs and Sr, respectively. From the long-term ISO leaching test, the 900day-leached fraction of Cs and Sr predicted by a semi-empirical model were 0.89% and 0.39%. The leaching behaviors indicated that Cs would be immobilized into a Si-rich phase while Sr would be in a P-rich phase. The experimental results revealed that the GRSS method could be an alternative method for a solidification of radioactive molten salt wastes.

Current Status of the Radioactive Waste Management Programme in Spain

Since 1984, ENRESA is responsible of the radioactive waste management and the decommissioning of nuclear installations in Spain. The major recent challenge has been the approval of the Sixth General Radioactive Waste Plan (GRWP) as “master plan” of the activities to be performed by ENRESA. Regarding the LILW programme, the El Cabril LILW disposal facility will be described highlighting the most relevant events especially focused on optimizing the existing capacity and the start-up of a purpose–built disposal area for VLLW. Concerning the HLW programme, two aspects may be distinguished in the direct management of spent fuel: temporary storage and long-term management. In this regards, a major challenge has been the decision adopted by the Spanish Government to set up a Interministerial Committee for the establishment of the criteria that must be met by the site of the Centralized Intermediate Storage (CTS) facility as the first and necessary step for the process. Also the developments of the long-term management programme will be presented in the frame of the ENRESA’s R&D programme. Finally, in the field of decommissioning they will be presented the PIMIC project at the CIEMAT centre and the activities in course for the decommissioning of Jose´ Cabrera NPP.

Lessons Learnt From Ignalina NPP Decommissioning Project

The Ignalina Nuclear Power Plant (INPP) is located in Lithuania, 130 km north of Vilnius, and consists of two 1500 MWe RBMK type units, commissioned respectively in December 1983 and August 1987. On the 1st of May 2004, the Republic of Lithuania became a member of the European Union. With the protocol on the Ignalina Nuclear Power in Lithuania which is annexed to the Accession Treaty, the Contracting Parties have agreed: - On Lithuanian side, to commit closure of unit 1 of INPP before 2005 and of Unit 2 by 31 December 2009; - On European Union side, to provide adequate additional Community assistance to the efforts of Lithuania to decommission INPP. The paper is divided in two parts. The first part describes how, starting from this agreement, the project was launched and organized, what is its present status and which activities are planned to reach the final ambitious objective of a green field. To give a global picture, the content of the different projects that were defined and the licensing process will also be presented. In the second part, the paper will focus on the lessons learnt. It will explain the difficulties encountered to define the decommissioning strategy, considering both immediate or differed dismantling options and why the first option was finally selected. The paper will mention other challenges and problems that the different actors of the project faced and how they were managed and solved. The paper will be written by representatives of the Ignalina NPP and of the Project Management Unit.

An Optimized Strategy for the Management of Spent Steam Generators

The decommissioning of the BR3 (Belgian Reactor 3) approaches its final phase. The electro-mechanical dismantling is almost completed and the program related to the decontamination of the building structures has been initiated. The issue of the evacuation of the primary circuit large components, and more specifically of the Steam Generator (SG), has been dealt successfully, applying innovative technologies to lead to remarkable results in terms of waste volume minimization and occupational radiation exposure. The strategy applied for the evacuation of the BR3 SG resulted from the elaboration and comparison of the following scenarios: • Closed loop chemical decontamination prior to dismantling, cutting and unconditional release or release after melting, • Cutting of the components without decontamination and evacuation of the materials in their respective waste categories, • Cutting, decontamination of the SG secondary side and evacuation of the full SG primary side to the melting facility for recycling. While the availability of the in-house developed MEDOC® process made the clearance of the SG bundle technically feasible, nuclear safety requirements and financial aspects were also in favour of the closed loop decontamination: minimization of contamination spreading and staff exposure during all subsequent manipulations, minimization of radwaste costs. For the segmentation of this component, different techniques have been considered: • An abrasive water jet (AWJ) cutting tool, • A prototype diamond wire developed for this application. The diamond wire allowed to cut in a single pass both the carbon steel shell and the stainless steel tube bundle. While the implementation of the diamond wire saw is rather simple, working conditions must be optimised to limit wearing of the wire and secondary waste production. Existing experience can be extrapolated to different legal frameworks in order to propose a financially and technically optimised “all-in” strategy for the management of “spent” SG.

Retrospection of Chernobyl Nuclear Accident for Decision Analysis Concerning Remedial Actions in Ukraine

It is considered the efficacy of decisions concerning remedial actions when of-site radiological monitoring in the early and (or) in the intermediate phases was absent or was not informative. There are examples of such situations in the former Soviet Union where many people have been exposed: releases of radioactive materials from “Krasnoyarsk-26” into Enisey River, releases of radioactive materials from “Chelabinsk-65” (the Kishtim accident), nuclear tests at the Semipalatinsk Test Site, the Chernobyl nuclear accident etc. If monitoring in the early and (or) in the intermediate phases is absent the decisions concerning remedial actions are usually developed on the base of permanent monitoring. However decisions of this kind may be essentially erroneous. For these cases it is proposed to make retrospection of radiological data of the early and intermediate phases of nuclear accident and to project decisions concerning remedial actions on the base of both retrospective data and permanent monitoring data. In this Report the indicated problem is considered by the example of the Chernobyl accident for Ukraine. Their of-site radiological monitoring in the early and intermediate phases was unsatisfactory. In particular, the pasture-cow-milk monitoring had not been made. All official decisions concerning dose estimations had been made on the base of measurements of 137Cs in body (40 measurements in 135 days and 55 measurements in 229 days after the Chernobyl accident). For the retrospection of radiological data of the Chernobyl accident dynamic model has been developed. This model has structure similar to the structure of Pathway model and Farmland model. Parameters of the developed model have been identified for agricultural conditions of Russia and Ukraine. By means of this model dynamics of 20 radionuclides in pathways and dynamics of doses have been estimated for the early, intermediate and late phases of the Chernobyl accident. The main results are following: • During the first year after the Chernobyl accident 75–93% of Commitment Effective Dose had been formed. • During the first year after the Chernobyl accident 85–90% of damage from radiation exposure had been formed. During the next 50 years (the late phase of accident) only 10–15% of damage from radiation exposure will have been formed. • Remedial actions (agricultural remedial actions as most effective) in Ukraine are intended for reduction of the damage from consumption of production which is contaminated in the late phase of accident. I.e. agricultural remedial actions have been intended for minimization only 10% of the total damage from radiation exposure. • Medical countermeasures can minimize radiation exposure damage by an order of magnitude greater than agricultural countermeasures. • Thus, retrospection of nuclear accident has essentially changed type of remedial actions and has given a chance to increase effectiveness of spending by an order of magnitude. This example illustrates that in order to optimize remedial actions it is required to use data of retrospection of nuclear accidents in all cases when monitoring in the early and (or) intermediate phases is unsatisfactory.

Flow and Radionuclide Transport From Rock to Surface Systems: Characterization and Modelling of Potential Repository Sites in Sweden

The safety assessments of potential geological repositories for spent nuclear fuel in Sweden are supported by modelling of groundwater flow in rock, to predict locations (exit points) where radionuclides from the deep repository may enter land, surface waters and associated ecosystems above the rock. This modelling includes detailed rock descriptions, but simplifies the upper part of the flow domain, including representations of meteorological processes and interactions with hydrological objects at the surface. Using the Laxemar candidate site as example, this paper investigates some potentially important consequences of these simplifications. Specifically, it compares particle tracking results obtained by a deep-rock groundwater flow model (CONNECTFLOW) and by MIKE SHE-MIKE 11, which contains detailed descriptions of near-surface/surface water flow. Overall, the models predict similar exit point patterns, occurring as clusters along streams in valleys, at a lake, and in sea bays. However, on a detailed level there are some prediction differences, which may be of importance for biosphere-focused safety assessments. CONNECTFLOW essentially predicts flow paths through the repository that follow fractures and deformation zones, outcropping in valleys. In comparison, MIKE SHE-MIKE 11 provides more detailed information on near-surface water flow paths, including the associated exit points and inputs to assessments of radionuclide retention.

A Heating Experiment in the Argillites in the Meuse/Haute-Marne Underground Research Laboratory

A heating experiment named TER is being conducted with the objectives to identify the thermal properties, as well as to enhance the knowledge on THM processes in the Callovo-Oxfordian clay at the Meuse/Haute Marne Underground Research Laboratory (France). The in situ experiment has being switched on from early 2006. The heater, 3 m length, is designed to inject the power in the undisturbed zone at 6 m from the gallery wall. A heater packer is inflated in a metallic tubing. During the experiment, numerous sensors are emplaced in the surrounding rock and are experienced to monitor the evolution in temperature, pore-water pressure and deformation. The models and numerical codes applied should be validated by comparing the modeling results with the measurements. In parallel, some lab testing have been achieved in order to compare the results given with two different scales (cm up to meter scale). In this paper, we present a general description of the TER experiment with installation of the heater equipment and the surrounding instrumentation. Details of the in situ measurements of temperature, pore-pressure and strain evolutions are given for the several heating and cooling phases. The thermal conductivity and some predominant parameters in THM processes (as linear thermal expansion coefficient and permeability) will be discussed.

The Integrated Project NF-PRO: Recent Advances in European Research on the Near-Field System

The Integrated Project NF-PRO (Sixth Framework Programme by the European Commission) investigates key-processes in the near-field of a geological repository for the disposal of high-level vitrified waste and spent fuel. The paper discusses the project scope and content and gives a summary overview of advances made by NF-PRO.

The Nuclear Fuel Services, Inc. Program to Support Disposition of Enriched Uranium-Bearing Materials

The disposition of surplus nuclear materials has become one of the most pressing issues of our time [1, 2]. Numerous agencies have invoked programs with the purpose of removing such materials from various international venues and dispositioning these materials in a manner that achieves non-proliferability. This paper describes the Nuclear Fuel Services, Inc (NFS) Nuclear Material Disposition Program, which to date has focused on a variety of Special Nuclear Material (SNM), in particular uranium of various enrichments. The major components of this program are discussed, with emphasis on recycle and return of material to the nuclear fuel cycle.